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Sample records for pebble springs-1 reactor

  1. PEBBLES: A COMPUTER CODE FOR MODELING PACKING, FLOW AND RECIRCULATIONOF PEBBLES IN A PEBBLE BED REACTOR

    SciTech Connect

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-10-01

    A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.

  2. Pebble Bed Reactor Dust Production Model

    SciTech Connect

    Abderrafi M. Ougouag; Joshua J. Cogliati

    2008-09-01

    The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

  3. Multiscale Analysis of Pebble Bed Reactors

    SciTech Connect

    Hans Gougar; Woo Yoon; Abderrafi Ougouag

    2010-10-01

    – The PEBBED code was developed at the Idaho National Laboratory for design and analysis of pebble-bed high temperature reactors. The diffusion-depletion-pebble-mixing algorithm of the original PEBBED code was enhanced through coupling with the THERMIX-KONVEK code for thermal fluid analysis and by the COMBINE code for online cross section generation. The COMBINE code solves the B-1 or B-3 approximations to the transport equation for neutron slowing down and resonance interactions in a homogeneous medium with simple corrections for shadowing and thermal self-shielding. The number densities of materials within specified regions of the core are averaged and transferred to COMBINE from PEBBED for updating during the burnup iteration. The simple treatment of self-shielding in previous versions of COMBINE led to inaccurate results for cross sections and unsatisfactory core performance calculations. A new version of COMBINE has been developed that treats all levels of heterogeneity using the 1D transport code ANISN. In a 3-stage calculation, slowing down is performed in 167 groups for each homogeneous subregion (kernel, particle layers, graphite shell, control rod absorber annulus, etc.) Particles in a local average pebble are homogenized using ANISN then passed to the next (pebble) stage. A 1D transport solution is again performed over the pebble geometry and the homogenized pebble cross sections are passed to a 1-d radial model of a wedge of the pebble bed core. This wedge may also include homogeneous reflector regions and a control rod region composed of annuli of different absorbing regions. Radial leakage effects are therefore captured with discrete ordinates transport while axial and azimuthal effects are captured with a transverse buckling term. In this paper, results of various PBR models will be compared with comparable models from literature. Performance of the code will be assessed.

  4. Experimental and computational investigation of flow of pebbles in a pebble bed nuclear reactor

    NASA Astrophysics Data System (ADS)

    Khane, Vaibhav B.

    The Pebble Bed Reactor (PBR) is a 4th generation nuclear reactor which is conceptually similar to moving bed reactors used in the chemical and petrochemical industries. In a PBR core, nuclear fuel in the form of pebbles moves slowly under the influence of gravity. Due to the dynamic nature of the core, a thorough understanding about slow and dense granular flow of pebbles is required from both a reactor safety and performance evaluation point of view. In this dissertation, a new integrated experimental and computational study of granular flow in a PBR has been performed. Continuous pebble re-circulation experimental set-up, mimicking flow of pebbles in a PBR, is designed and developed. Experimental investigation of the flow of pebbles in a mimicked test reactor was carried out for the first time using non-invasive radioactive particle tracking (RPT) and residence time distribution (RTD) techniques to measure the pebble trajectory, velocity, overall/zonal residence times, flow patterns etc. The tracer trajectory length and overall/zonal residence time is found to increase with change in pebble's initial seeding position from the center towards the wall of the test reactor. Overall and zonal average velocities of pebbles are found to decrease from the center towards the wall. Discrete element method (DEM) based simulations of test reactor geometry were also carried out using commercial code EDEM(TM) and simulation results were validated using the obtained benchmark experimental data. In addition, EDEM(TM) based parametric sensitivity study of interaction properties was carried out which suggests that static friction characteristics play an important role from a packed/pebble beds structural characterization point of view. To make the RPT technique viable for practical applications and to enhance its accuracy, a novel and dynamic technique for RPT calibration was designed and developed. Preliminary feasibility results suggest that it can be implemented as a non

  5. The effects of temperatures on the pebble flow in a pebble bed high temperature reactor

    SciTech Connect

    Sen, R. S.; Cogliati, J. J.; Gougar, H. D.

    2012-07-01

    The core of a pebble bed high temperature reactor (PBHTR) moves during operation, a feature which leads to better fuel economy (online refueling with no burnable poisons) and lower fuel stress. The pebbles are loaded at the top and trickle to the bottom of the core after which the burnup of each is measured. The pebbles that are not fully burned are recirculated through the core until the target burnup is achieved. The flow pattern of the pebbles through the core is of importance for core simulations because it couples the burnup distribution to the core temperature and power profiles, especially in cores with two or more radial burnup 'zones '. The pebble velocity profile is a strong function of the core geometry and the friction between the pebbles and the surrounding structures (other pebbles or graphite reflector blocks). The friction coefficient for graphite in a helium environment is inversely related to the temperature. The Thorium High Temperature Reactor (THTR) operated in Germany between 1983 and 1989. It featured a two-zone core, an inner core (IC) and outer core (OC), with different fuel mixtures loaded in each zone. The rate at which the IC was refueled relative to the OC in THTR was designed to be 0.56. During its operation, however, this ratio was measured to be 0.76, suggesting the pebbles in the inner core traveled faster than expected. It has been postulated that the positive feedback effect between inner core temperature, burnup, and pebble flow was underestimated in THTR. Because of the power shape, the center of the core in a typical cylindrical PBHTR operates at a higher temperature than the region next to the side reflector. The friction between pebbles in the IC is lower than that in the OC, perhaps causing a higher relative flow rate and lower average burnup, which in turn yield a higher local power density. Furthermore, the pebbles in the center region have higher velocities than the pebbles next to the side reflector due to the

  6. PEBBLES

    SciTech Connect

    Cogliati, Joshua J.

    2010-09-01

    The PEBBLES code is a computer program designed to simulate the motion, packing and vibration of spheres that undergo various mechanical forces including gravitation, Hooke's law force and various friction forces. The frictional forces include true static friction that allows non-zero angles of repose. Each pebble is individually simulated using the distinct element method. The program outputs various tallies as textual numbers. These tallies include pebble position, pebble angular and linear velocity, force on the wall and between pebbles, probabilities of pebbles moving between different locations, accumulated amount of linear motion between pebbles, and average velocity in different regions of the container.

  7. PEBBLES

    2010-09-01

    The PEBBLES code is a computer program designed to simulate the motion, packing and vibration of spheres that undergo various mechanical forces including gravitation, Hooke's law force and various friction forces. The frictional forces include true static friction that allows non-zero angles of repose. Each pebble is individually simulated using the distinct element method. The program outputs various tallies as textual numbers. These tallies include pebble position, pebble angular and linear velocity, force on themore » wall and between pebbles, probabilities of pebbles moving between different locations, accumulated amount of linear motion between pebbles, and average velocity in different regions of the container.« less

  8. PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

    SciTech Connect

    Vondy, D.R.

    1981-09-01

    This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity. The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.

  9. Pebble Bed Reactor review update. Fiscal year 1979 annual report

    SciTech Connect

    Not Available

    1980-01-01

    Updated information is presented on the Pebble Bed Reactor (PBR) concept being developed in the Federal Republic of Germany for electricity generation and process heat applications. Information is presented concerning nuclear analysis and core performance, fuel cycle evaluation, reactor internals, and safety and availability.

  10. PEBBED ANALYSIS OF HOT SPOTS IN PEBBLE-BED REACTORS

    SciTech Connect

    Abderrafi M. Ougouag; Hans D. Gougar; William K. Terry; Frederik Reitsma; Wessel Joubert

    2005-09-01

    The Idaho National Laboratory’s PEBBED code and simple probability considerations are used to estimate the likelihood and consequences of the accumulation of highly reactive pebbles in the region of peak power in a pebble-bed reactor. The PEBBED code is briefly described, and the logic of the probability calculations is presented in detail. The results of the calculations appear to show that hot-spot formation produces only moderate increases in peak accident temperatures, and no increases at all in normal operating temperatures.

  11. Automated Design and Optimization of Pebble-bed Reactor Cores

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2010-07-01

    We present a conceptual design approach for high-temperature gas-cooled reactors using recirculating pebble-bed cores. The design approach employs PEBBED, a reactor physics code specifically designed to solve for and analyze the asymptotic burnup state of pebble-bed reactors, in conjunction with a genetic algorithm to obtain a core that maximizes a fitness value that is a function of user-specified parameters. The uniqueness of the asymptotic core state and the small number of independent parameters that define it suggest that core geometry and fuel cycle can be efficiently optimized toward a specified objective. PEBBED exploits a novel representation of the distribution of pebbles that enables efficient coupling of the burnup and neutron diffusion solvers. With this method, even complex pebble recirculation schemes can be expressed in terms of a few parameters that are amenable to modern optimization techniques. With PEBBED, the user chooses the type and range of core physics parameters that represent the design space. A set of traits, each with acceptable and preferred values expressed by a simple fitness function, is used to evaluate the candidate reactor cores. The stochastic search algorithm automatically drives the generation of core parameters toward the optimal core as defined by the user. The optimized design can then be modeled and analyzed in greater detail using higher resolution and more computationally demanding tools to confirm the desired characteristics. For this study, the design of pebble-bed high temperature reactor concepts subjected to demanding physical constraints demonstrated the efficacy of the PEBBED algorithm.

  12. Core Optimization of a Deep-Burn Pebble Bed Reactor

    SciTech Connect

    Brian Boer; Abderrafi M. Ougouag

    2010-06-01

    Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).

  13. Tightly Coupled Multiphysics Algorithm for Pebble Bed Reactors

    SciTech Connect

    HyeongKae Park; Dana Knoll; Derek Gaston; Richard Martineau

    2010-10-01

    We have developed a tightly coupled multiphysics simulation tool for the pebble-bed reactor (PBR) concept, a type of Very High-Temperature gas-cooled Reactor (VHTR). The simulation tool, PRONGHORN, takes advantages of the Multiphysics Object-Oriented Simulation Environment library, and is capable of solving multidimensional thermal-fluid and neutronics problems implicitly with a Newton-based approach. Expensive Jacobian matrix formation is alleviated via the Jacobian-free Newton-Krylov method, and physics-based preconditioning is applied to minimize Krylov iterations. Motivation for the work is provided via analysis and numerical experiments on simpler multiphysics reactor models. We then provide detail of the physical models and numerical methods in PRONGHORN. Finally, PRONGHORN's algorithmic capability is demonstrated on a number of PBR test cases.

  14. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  15. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  16. Proliferation resistant fuel for pebble bed modular reactors

    SciTech Connect

    Ronen, Y.; Aboudy, M.; Regev, D.; Gilad, E.

    2012-07-01

    We show that it is possible to denature the Plutonium produced in Pebble Bed Modular Reactors (PBMR) by doping the nuclear fuel with either 3050 ppm of {sup 237}Np or 2100 ppm of Am vector. A correct choice of these isotopes concentration yields denatured Plutonium with isotopic ratio {sup 238}Pu/Pu {>=} 6%, for the entire fuel burnup cycle. The penalty for introducing these isotopes into the nuclear fuel is a subsequent shortening of the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full Power Days (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and 1965 MWd/ton reduction in fuel discharge burnup. (authors)

  17. Calculational approach and results of the safe shutdown earthquake event for the pebble bed modular reactor

    SciTech Connect

    Van Heerden, G.; Sen, S.; Reitsma, F.

    2006-07-01

    The Pebble Bed Modular Reactor (PBMR) concept can be described as a high-temperature helium-cooled, graphite-moderated pebble-bed reactor with a multi-pass fuelling scheme. The fuel is contained in 6 cm diameter graphite spheres containing carbon-based coated UO{sub 2} kernels. An online fuel reload scheme is applied with the fuel spheres being circulated through the reactor. The pebble-bed reactor core thus consists of fuel pebbles packed in the core cavity in a random way. The packing densities and pebble flow is well known through analysis and tests done in the German experimental and development program. The pebble-bed typically has a packing fraction of 0.61. In the event of an earthquake this packing fraction may increase with the effect that the core geometry and core reactivity will change. The Safe Shutdown Earthquake (SSE) analysis performed for the PBMR 400 MW design is described in this paper, and it specifically covers SSE-induced pebble-bed packing fractions of 0.62 and 0.64. The main effects governing the addition of reactivity in the SSE event are the changes in core neutronic leakage due to the decreased core size and the decreased effectiveness of the control rods as the pebble-bed height decreases. This paper describes the models, methods and tools used to analyse the event, the results obtained for the different approaches and the consequences and safety implications of such an event. (authors)

  18. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    SciTech Connect

    Mcwilliams, A. J.

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  19. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    SciTech Connect

    Peterson, Per; Greenspan, Ehud

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  20. Modular Pebble Bed Reactor Project, University Research Consortium Annual Report

    SciTech Connect

    Petti, David Andrew

    2000-07-01

    This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: · Improved fuel particle performance · Reactor physics · Economics · Proliferation resistance · Power conversion system modeling · Safety analysis · Regulatory and licensing strategy Recent accomplishments include: · Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. · A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. · Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. · A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a master’s degree thesis. · Safety issues associated with air ingress are being evaluated. · Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. · PEBBED, a fast deterministic neutronic code package suitable for

  1. METHODS FOR MODELING THE PACKING OF FUEL ELEMENTS IN PEBBLE BED REACTORS

    SciTech Connect

    Abderrafi M. Ougouag; Joshua J. Cogliati; Jan-Leen Kloosterman

    2005-09-01

    Two methods for the modeling of the packing of pebbles in the pebble bed reactors are presented and compared. The first method is based on random generation of potential centers for the pebbles, followed by rejection of points that are not compatible with the geometric constraint of no (or limited) pebbles overlap. The second method models the actual physical packing process, accounting for the dynamic of pebbles as they are dropped onto the pebble bed and as they settle therein. A simplification in the latter model is the assumption of a starting point with very dilute packing followed by settling. The results from the two models are compared and the properties of the second model and the dependence of its results on many of the modeling parameters are presented. The first model (with no overlap allowed) has been implemented into a code to compute Dancoff factors. The second model will soon be implemented into that same code and will also be used to model flow of pebbles in a reactor and core densification in the simulation of earthquakes. Both methods reproduce experimental values well, with the latter displaying a high level of fidelity.

  2. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  3. Computational and experimental prediction of dust production in pebble bed reactors, Part II

    SciTech Connect

    Mie Hiruta; Gannon Johnson; Maziar Rostamian; Gabriel P. Potirniche; Abderrafi M. Ougouag; Massimo Bertino; Louis Franzel; Akira Tokuhiro

    2013-10-01

    This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor.

  4. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    SciTech Connect

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  5. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  6. Analysis of granular flow in a pebble-bed nuclear reactor

    SciTech Connect

    Rycroft, C H; Grest, Gary S; Landry, James W; Bazant, Martin Z

    2006-04-17

    Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a ma jor impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440,000 frictional, viscoelastic 6cm-diameter spheres draining in a cylindrical vessel of diameter 3.5m and height 10m with bottom funnels angled at 30° or 60° . We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

  7. Analysis of granular flow in a pebble-bed nuclear reactor.

    PubMed

    Rycroft, Chris H; Grest, Gary S; Landry, James W; Bazant, Martin Z

    2006-08-01

    Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a major impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440,000 frictional, viscoelastic 6-cm-diam spheres draining in a cylindrical vessel of diameter 3.5m and height 10 m with bottom funnels angled at 30 degrees or 60 degrees. We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

  8. Deleterious Thermal Effects Due To Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

    NASA Technical Reports Server (NTRS)

    Moran, Robert P.

    2013-01-01

    A review of literature associated with Pebble Bed and Particle Bed reactor core research has revealed a systemic problem inherent to reactor core concepts which utilize randomized rather than structured coolant channel flow paths. For both the Pebble Bed and Particle Bed Reactor designs; case studies reveal that for indeterminate reasons, regions within the core would suffer from excessive heating leading to thermal runaway and localized fuel melting. A thermal Computational Fluid Dynamics model was utilized to verify that In both the Pebble Bed and Particle Bed Reactor concepts randomized coolant channel pathways combined with localized high temperature regions would work together to resist the flow of coolant diverting it away from where it is needed the most to cooler less resistive pathways where it is needed the least. In other words given the choice via randomized coolant pathways the reactor coolant will take the path of least resistance, and hot zones offer the highest resistance. Having identified the relationship between randomized coolant channel pathways and localized fuel melting it is now safe to assume that other reactor concepts that utilize randomized coolant pathways such as the foam core reactor are also susceptible to this phenomenon.

  9. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  10. Stress Analysis of Coated Particle Fuel in the Deep-Burn Pebble Bed Reactor Design

    SciTech Connect

    B. Boer; A. M. Ougouag

    2010-05-01

    High fuel temperatures and resulting fuel particle coating stresses can be expected in a Pu and minor actinide fueled Pebble Bed Modular Reactor (400 MWth) design as compared to the ’standard’ UO2 fueled core. The high discharge burnup aimed for in this Deep-Burn design results in increased power and temperature peaking in the pebble bed near the inner and outer reflector. Furthermore, the pebble power in a multi-pass in-core pebble recycling scheme is relatively high for pebbles that make their first core pass. This might result in an increase of the mechanical failure of the coatings, which serve as the containment of radioactive fission products in the PBMR design. To investigate the integrity of the particle fuel coatings as a function of the irradiation time (i.e. burnup), core position and during a Loss Of Forced Cooling (LOFC) incident the PArticle STress Analysis code (PASTA) has been coupled to the PEBBED code for neutronics, thermal-hydraulics and depletion analysis of the core. Two deep burn fuel types (Pu with or without initial MA fuel content) have been investigated with the new code system for normal and transient conditions including the effect of the statistical variation of thickness of the coating layers.

  11. Feasibility of Burning First- and Second-Generation Plutonium in Pebble Bed High-Temperature Reactors

    SciTech Connect

    Haas, J.B.M. de; Kuijper, J.C

    2005-08-15

    The core physics investigations at the Nuclear Research Consultancy Group in the Netherlands, as part of the activities within the HTR-N project of the European Fifth Framework Program, are focused on the incineration of pure (first- and second-generation) Pu fuels in the reference pebble bed high-temperature gas-cooled reactor (HTR) HTR-MODUL with a continuous reload [MEDUL, (MEhrfach DUrchLauf, multipass)] fueling strategy in which the spherical fuel elements, or pebbles, pass through the core a number of times before being permanently discharged. For pebbles fueled with different loadings of plutonium, the feasibility of a sustained fuel cycle under nominal reactor conditions was investigated by means of the reactivity and temperature coefficients of the reactor. The HTR-MODUL was found to be a very effective reactor to reduce the stockpile of first-generation plutonium. It reduces the amount of plutonium to about one-sixth of the original and reduces the risk of proliferation by denaturing the plutonium vector. For second-generation plutonium the incineration is less favorable, as the amount of plutonium is only halved.

  12. On-line interrogation of pebble bed reactor fuel using passive gamma-ray spectrometry

    NASA Astrophysics Data System (ADS)

    Chen, Jianwei

    The Pebble Bed Reactor (PBR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor. In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles (of varying enrichment) are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burnup limit (˜80,000--100,000 MWD/MTU). Unlike the situation with conventional light water reactors (LWRs), depending solely on computational methods to perform in-core fuel management will be highly inaccurate. As a result, an on-line measurement approach becomes the only accurate method to assess whether a particular pebble has reached its end-of-life burnup limit. In this work, an investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line interrogation of PBR fuel for the simultaneous determination of burnup and enrichment on a pebble-by-pebble basis. Due to the unavailability of irradiated or fresh pebbles, Monte Carlo simulations were used to study the gamma-ray spectra of the PBR fuel at various levels of burnup. A pebble depletion calculation was performed using the ORIGEN code, which yielded the gamma-ray source term that was introduced into the input of an MCNP simulation. The MCNP simulation assumed the use of a high-purity coaxial germanium detector. Due to the lack of one-group high temperature reactor cross sections for ORIGEN, a heterogeneous MCNP model was developed to describe a typical PBR core. Subsequently, the code MONTEBURNS was used to couple the MCNP model and ORIGEN. This approach allowed the development of the burnup-dependent, one-group spectral-averaged PBR cross sections to be used in the ORIGEN pebble depletion calculation. Based on the above studies, a relative approach for performing the measurements was established. The approach is based on using the relative activities of Np-239/I-132 in combination

  13. Pebble bed reactor fiscal year 1980: review summary report

    SciTech Connect

    Not Available

    1980-07-01

    Information on high-temperature reactor development is presented concerning reactor operating experience; core performance assessment; core control and shutdown; reflector and core support; maintenance and availability; safety aspects of PBR and prismatic comparison; PCRV dimensions; and fuel reprocessing cost estimate.

  14. A simplified DEM-CFD approach for pebble bed reactor simulations

    SciTech Connect

    Li, Y.; Ji, W.

    2012-07-01

    In pebble bed reactors (PBR's), the pebble flow and the coolant flow are coupled with each other through coolant-pebble interactions. Approaches with different fidelities have been proposed to simulate similar phenomena. Coupled Discrete Element Method-Computational Fluid Dynamics (DEM-CFD) approaches are widely studied and applied in these problems due to its good balance between efficiency and accuracy. In this work, based on the symmetry of the PBR geometry, a simplified 3D-DEM/2D-CFD approach is proposed to speed up the DEM-CFD simulation without significant loss of accuracy. Pebble flow is simulated by a full 3-D DEM, while the coolant flow field is calculated with a 2-D CFD simulation by averaging variables along the annular direction in the cylindrical geometry. Results show that this simplification can greatly enhance the efficiency for cylindrical core, which enables further inclusion of other physics such as thermal and neutronic effect in the multi-physics simulations for PBR's. (authors)

  15. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    SciTech Connect

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  16. Safeguards Challenges for Pebble-Bed Reactors (PBRs):Peoples Republic of China (PRC)

    SciTech Connect

    Forsberg, Charles W.; Moses, David Lewis

    2009-11-01

    The Peoples Republic of China (PRC) is operating the HTR-10 pebble-bed reactor (PBR) and is in the process of building a prototype PBR plant with two modular reactors (250-MW(t) per reactor) feeding steam to a single turbine-generator. It is likely to be the first modular hightemperature reactor to be ready for commercial deployment in the world because it is a highpriority project for the PRC. The plant design features multiple modular reactors feeding steam to a single turbine generator where the number of modules determines the plant output. The design and commercialization strategy are based on PRC strengths: (1) a rapidly growing electric market that will support low-cost mass production of modular reactor units and (2) a balance of plant system based on economics of scale that uses the same mass-produced turbine-generator systems used in PRC coal plants. If successful, in addition to supplying the PRC market, this strategy could enable China to be the leading exporter of nuclear reactors to developing countries. The modular characteristics of the reactor match much of the need elsewhere in the world. PBRs have major safety advantages and a radically different fuel. The fuel, not the plant systems, is the primary safety system to prevent and mitigate the release of radionuclides under accident conditions. The fuel consists of small (6-cm) pebbles (spheres) containing coatedparticle fuel in a graphitized carbon matrix. The fuel loading per pebble is small (~9 grams of low-enriched uranium) and hundreds of thousands of pebbles are required to fuel a nuclear plant. The uranium concentration in the fuel is an order of magnitude less than in traditional nuclear fuels. These characteristics make the fuel significantly less attractive for illicit use (weapons production or dirty bomb); but, its unusual physical form may require changes in the tools used for safeguards. This report describes PBRs, what is different, and the safeguards challenges. A series of

  17. Plutonium and minor actinide utilisation in a pebble-bed high temperature reactor

    SciTech Connect

    Petrov, B. Y.; Kuijper, J. C.; Oppe, J.; De Haas, J. B. M.

    2012-07-01

    This paper contains results of the analysis of the pebble-bed high temperature gas-cooled PUMA reactor loaded with plutonium and minor actinide (Pu/MA) fuel. Starting from knowledge and experience gained in the Euratom FP5 projects HTR-N and HTR-N1, this study aims at demonstrating the potential of high temperature reactors to utilize or transmute Pu/MA fuel. The work has been performed within the Euratom FP6 project PUMA. A number of different fuel types and fuel configurations have been analyzed and compared with respect to incineration performance and safety-related reactor parameters. The results show the excellent plutonium and minor actinide burning capabilities of the high temperature reactor. The largest degree of incineration is attained in the case of an HTR fuelled by pure plutonium fuel as it remains critical at very deep burnup of the discharged pebbles. Addition of minor actinides to the fuel leads to decrease of the achievable discharge burnup and therefore smaller fraction of actinides incinerated during reactor operation. The inert-matrix fuel design improves the transmutation performance of the reactor, while the 'wallpaper' fuel does not have advantage over the standard fuel design in this respect. After 100 years of decay following the fuel discharge, the total amount of actinides remains almost unchanged for all of the fuel types considered. Among the plutonium isotopes, only the amount of Pu-241 is reduced significantly due to its relatively short half-life. (authors)

  18. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    SciTech Connect

    Beyer, Brian David; Beddingfield, David H; Durst, Philip; Bean, Robert

    2010-01-01

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

  19. Supplemental Report on Nuclear Safeguards Considerations for the Pebble Bed Modular Reactor (PBMR)

    SciTech Connect

    Moses, David Lewis; Ehinger, Michael H

    2010-05-01

    Recent reports by Department of Energy National Laboratories have discussed safeguards considerations for the low enriched uranium (LEU) fueled Pebble Bed Modular Reactor (PBMR) and the need for bulk accountancy of the plutonium in used fuel. These reports fail to account effectively for the degree of plutonium dilution in the graphitized-carbon pebbles that is sufficient to meet the International Atomic Energy Agency's (IAEA's) 'provisional' guidelines for termination of safeguards on 'measured discards.' The thrust of this finding is not to terminate safeguards but to limit the need for specific accountancy of plutonium in stored used fuel. While the residual uranium in the used fuel may not be judged sufficiently diluted to meet the IAEA provisional guidelines for termination of safeguards, the estimated quantities of {sup 232}U and {sup 236}U in the used fuel at the target burn-up of {approx}91 GWD/MT exceed specification limits for reprocessed uranium (ASTM C787) and will require extensive blending with either natural uranium or uranium enrichment tails to dilute the {sup 236}U content to fall within specification thus making the PBMR used fuel less desirable for commercial reprocessing and reuse than that from light water reactors. Also the PBMR specific activity of reprocessed uranium isotopic mixture and its A{sub 2} values for effective dose limit if released in a dispersible form during a transportation accident are more limiting than the equivalent values for light water reactor spent fuel at 55 GWD/MT without accounting for the presence of the principal carry-over fission product ({sup 99}Tc) and any possible plutonium contamination that may be present from attempted covert reprocessing. Thus, the potentially recoverable uranium from PBMR used fuel carries reactivity penalties and radiological penalties likely greater than those for reprocessed uranium from light water reactors. These factors impact the economics of reprocessing, but a more significant

  20. Pebble bed pebble motion: Simulation and Application

    NASA Astrophysics Data System (ADS)

    Cogliati, Joshua J.

    Pebble bed reactors (PBR) have moving graphite fuel pebbles. This unique feature provides advantages, but also means that simulation of the reactor requires understanding the typical motion and location of the granular flow of pebbles. This dissertation presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, the PEBBLES code has been parallelized. PEBBLES runs on shared memory architectures and distributed memory architectures. For the shared memory architecture, the code uses a new O(n) lock-less parallel collision detection algorithm to determine which pebbles are likely to be in contact. The new collision detection algorithm improves on the traditional non-parallel O(n log(n)) collision detection algorithm. These features combine to form a fast parallel pebble motion simulation. The PEBBLES code provides new capabilities for understanding and optimizing PBRs. The PEBBLES code has provided the pebble motion data required to calculate the motion of pebbles during a simulated earthquake. The PEBBLES code provides the ability to

  1. The preliminary analysis on the steady-state and kinetic features of the molten salt pebble-bed reactor

    SciTech Connect

    Xia, B.; Lu, Y.

    2012-07-01

    A novel design concept of molten salt pebble-bed reactor with an ultra-simplified integral primary circuit called 'Nuclear Hot Spring' has been proposed, featured by horizontal coolant flow in a deep pool pebble-bed reactor, providing 'natural safety' features with natural circulation under full power operation and less expensive primary circuit arrangement. In this work, the steady-state physical properties of the equilibrium state of the molten salt pebble-bed reactor are calculated by using the VSOP code, and the steady-state thermo-hydraulic analysis is carried out based on the approximation of absolutely horizontal flow of the coolant through the core. A new concept of 2-dimensional, both axial and radial, multi-pass on-line fuelling scheme is presented. The result reveals that the radial multi-pass scheme provides more flattened power distribution and safer temperature distribution than the one-pass scheme. A parametric analysis is made corresponding to different pebble diameters, the key parameter of the core resistance and the temperature at the pebble center. It is verified that within a wide range of pebble diameters, the maximum pebble center temperatures are far below the safety limit of the fuel, and the core resistance is considerably less than the buoyant force, indicating that the natural circulation under full power operation is achievable and the ultra-simplified integral primary circuit without any pump is possible. For the kinetic properties, it is verified that the negative temperature coefficient is achieved in sufficient under-moderated condition through the preliminary analysis on the temperature coefficients of fuel, coolant and moderator. The requirement of reactivity compensation at the shutdown stages of the operation period is calculated for the further studies on the reactivity control. The molten salt pebble-bed reactor with horizontal coolant flow can provide enhanced safety and economical features. (authors)

  2. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

    SciTech Connect

    H. D. Gougar; C. B. Davis

    2006-04-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

  3. PEBBLES Mechanics Simulation Speedup

    SciTech Connect

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2010-05-01

    Pebble bed reactors contain large numbers of spherical fuel elements arranged randomly. Determining the motion and location of these fuel elements is required for calculating certain parameters of pebble bed reactor operation. These simulations involve hundreds of thousands of pebbles and involve determining the entire core motion as pebbles are recirculated. Single processor algorithms for this are insufficient since they would take decades to centuries of wall-clock time. This paper describes the process of parallelizing and speeding up the PEBBLES pebble mechanics simulation code. Both shared memory programming with the Open Multi-Processing API and distributed memory programming with the Message Passing Interface API are used in simultaneously in this process. A new shared memory lock-less linear time collision detection algorithm is described. This method allows faster detection of pebbles in contact than generic methods. These combine to make full recirculations on AVR sized reactors possible in months of wall clock time.

  4. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    SciTech Connect

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  5. On the evaluation of pebble bed reactor critical experiments using the PEBBED code

    SciTech Connect

    Hans D. Gougar; R. Sonat Sen

    2001-10-01

    The PEBBED pebble bed reactor fuel management code under development at the Idaho National Laboratory is designed for rapid design and analysis of pebble bed high temperature reactors (PBRs). Embedded within the code are the THERMIX-KONVEK thermal fluid solver and the COMBINE-7 spectrum generation code for inline cross section homogenization. Because 1D symmetry can be found at each stage of core heterogeneity; spherical at TRISO and pebble levels, and cylindrical at the control rod and core levels, the 1-D transport capability of ANISN is assumed to be sufficient in most cases for generating flux solutions for cross section homogenization. Furthermore, it is fast enough to be executed during the analysis or the equilibrium core. Multi-group diffusion-based design codes such as PEBBED and VSOP are not expected to yield the accuracy and resolution of continuous energy Monte Carlo codes for evaluation of critical experiments. Nonetheless, if the preparation of multigroup cross sections can adequately capture the physics of the mixing of PBR fuel elements and leakage from the core, reasonable results may be obtained. In this paper, results of the application of PEBBED to two critical experiments (HTR Proteus and HTR-10) and associated computational models are presented. The embedded 1-D transport solver is shown to capture the double heterogeneity of the pebble fuel in unit cell calculations. Eigenvalue calculations of a whole core are more challenging, particularly if the boron concentration is uncertain. The sensitivity of major safety parameters to variations in modeling assumptions, however, is shown to be minimal. The embedded transport solver can also be used to obtain control rod worths but only with adjustment of the local spectrum. Results are compared to those of other codes as well as Core 4 of the HTR-Proteus experiment which contains partially inserted rods. They indicate the need for a reference solution to adjust the radius of the graphite in the

  6. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    SciTech Connect

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  7. Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

    NASA Astrophysics Data System (ADS)

    Fratoni, Massimiliano

    This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be

  8. Analysis of the impact of random summing on passive assay of pebble bed reactor fuel using gamma-ray spectrometry

    NASA Astrophysics Data System (ADS)

    Chen, J.; Hawari, A. I.

    2007-08-01

    Pebble bed reactors (PBR) are characterized by multi-pass fuel systems in which spherical fuel pebbles are circulated through the core until they reach a proposed burnup limit. The fuel is assayed on-line to ensure that the burnup limit is not breached. However, random summing effects can impact the response of the burnup measurement system and result in distortions that degrade the accuracy of the assay results. Monte Carlo analysis was performed to estimate the magnitude and effect of random summing on the absolute and relative indicators that have been identified as usable in on-line assay. For a throughput rate of 10 5 counts/s and trapezoidal pulse shaping of the signals, the results show that absolute indicators suffer from severe distortions due to this effect. Relative indicators are found to be resistant to random summing with the deviation in the ratio of peak areas remaining less than 5-15% depending on pulse width.

  9. The correction of pebble bed reactor nodal cross sections for the effects of leakage and depletion history

    NASA Astrophysics Data System (ADS)

    Hudson, Nathanael Harrison

    An accurate and computationally fast method to generate nodal cross sections for the Pebble Bed Reactor (PBR) was presented. In this method, named Spectral History Correction (SHC), a set of fine group microscopic cross section libraries, pre-computed at specified depletion and moderation states, was coupled with the nodal nuclide densities and group bucklings to compute the new fine group spectrum for each node. The relevant fine group cross-section library was then recollapsed to the local broad group cross-section structure with this new fine group spectrum. This library set was tracked in terms of fuel isotopic densities. Fine group modulation factors (to correct the homogeneous flux for heterogeneous effects) and fission spectra were also stored with the cross section library. As the PBR simulation converges to a steady state fuel cycle, the initial nodal cross section library becomes inaccurate due to the burnup of the fuel and the neutron leakage into and out of the node. Because of the recirculation of discharged fuel pebbles with fresh fuel pebbles, a node can consist of a collection of pebbles at various burnup stages. To account for the nodal burnup, the microscopic cross sections were combined with nodal averaged atom densities to approximate the fine group macroscopic cross-sections for that node. These constructed, homogeneous macroscopic cross sections within the node were used to calculate a numerical solution for the fine group spectrum with B1 theory. This new fine spectrum was used to collapse the pre-computed microscopic cross section library to the broad group structure employed by the fuel cycle code. This SHC technique was developed and practically implemented as a subroutine within the PBR fuel cycle code PEBBED. The SHC subroutine was called to recalculate the broad group cross sections during the code convergence. The result was a fast method that compared favorably to the benchmark scheme of cross section calculation with the lattice

  10. Pebble-bed pebble motion: Simulation and Applications

    SciTech Connect

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2011-11-01

    Pebble bed reactors (PBR) have moving graphite fuel pebbles. This unique feature provides advantages, but also means that simulation of the reactor requires understanding the typical motion and location of the granular flow of pebbles. This report presents a method for simulation of motion of the pebbles in a PBR. A new mechanical motion simulator, PEBBLES, efficiently simulates the key elements of motion of the pebbles in a PBR. This model simulates gravitational force and contact forces including kinetic and true static friction. It's used for a variety of tasks including simulation of the effect of earthquakes on a PBR, calculation of packing fractions, Dancoff factors, pebble wear and the pebble force on the walls. The simulator includes a new differential static friction model for the varied geometries of PBRs. A new static friction benchmark was devised via analytically solving the mechanics equations to determine the minimum pebble-to-pebble friction and pebble-to-surface friction for a five pebble pyramid. This pyramid check as well as a comparison to the Janssen formula was used to test the new static friction equations. Because larger pebble bed simulations involve hundreds of thousands of pebbles and long periods of time, the PEBBLES code has been parallelized. PEBBLES runs on shared memory architectures and distributed memory architectures. For the shared memory architecture, the code uses a new O(n) lock-less parallel collision detection algorithm to determine which pebbles are likely to be in contact. The new collision detection algorithm improves on the traditional non-parallel O(n log(n)) collision detection algorithm. These features combine to form a fast parallel pebble motion simulation. The PEBBLES code provides new capabilities for understanding and optimizing PBRs. The PEBBLES code has provided the pebble motion data required to calculate the motion of pebbles during a simulated earthquake. The PEBBLES code provides the ability to determine

  11. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    NASA Astrophysics Data System (ADS)

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling

  12. Automated spectral zones selection methodology for diffusion theory data preparation for pebble bed reactor analysis

    NASA Astrophysics Data System (ADS)

    Mphahlele, Ramatsemela

    A methodology is developed for the determination of the optimum spectral zones in Pebble Bed Reactors (PBR). In this work a spectral zone is defined as a zone made up of a number of nodes whose characteristics are collectively similar and that are assigned the same few-group diffusion constants. In other words the spectral zones are the regions over which the few-group diffusion parameters are generated. The identification of spectral boundaries is treated as an optimization problem. It is solved by systematically and simultaneously repositioning all zone boundaries to achieve the global minimum error between the reference transport solution (MCNP) and the diffusion code solution (NEM). The objective function for the optimization algorithm is the total reaction rate error, which is defined as the sum of the leakage, absorption and fission reaction rates error in each zone. An iterative determination of group-dependent bucklings is incorporated into the methodology to properly account for spectral effects of neighboring zones. A preferred energy group structure has also been chosen. This optimization approach with the reference transport solution has proved to be accurate and consistent, however the computational effort required to complete the optimization process is significant. Thus a more practical methodology is also developed for the determination of the spectral zones in PBRs. The reactor physics characteristics of the spectral zones have been studied to understand the nature of the spectral zone boundaries. The practical tool involves the use of spectral indices based on few-group diffusion theory whole core calculations. With this methodology, there is no need to first have a reference transport solution. It is shown that the diffusion-theory coarse group fluxes and the effective multiplication factor computed using zones based on the practical index agrees within a narrow tolerance with those of the reference approach. Therefore the "practical" index

  13. Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Pebble Fuel

    SciTech Connect

    Philip Casey Durst; Mark Schanfein

    2012-08-01

    The following is a guidance document from a series prepared for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), under the Next Generation Safeguards Initiative (NGSI), to assist facility designers and operators in implementing international Safeguards-by-Design (SBD). SBD has two main objectives: (1) to avoid costly and time consuming redesign work or retrofits of new nuclear fuel cycle facilities and (2) to make the implementation of international safeguards more effective and efficient at such facilities. In the long term, the attainment of these goals would save industry and the International Atomic Energy Agency (IAEA) time, money, and resources and be mutually beneficial. This particular safeguards guidance document focuses on pebble fuel high temperature gas reactors (HTGR). The purpose of the IAEA safeguards system is to provide credible assurance to the international community that nuclear material and other specified items are not diverted from peaceful nuclear uses. The safeguards system consists of the IAEA’s statutory authority to establish safeguards; safeguards rights and obligations in safeguards agreements and additional protocols; and technical measures implemented pursuant to those agreements. Of foremost importance is the international safeguards agreement between the country and the IAEA, concluded pursuant to the Treaty on the Non-Proliferation of Nuclear Weapons (NPT). According to a 1992 IAEA Board of Governors decision, countries must: notify the IAEA of a decision to construct a new nuclear facility as soon as such decision is taken; provide design information on such facilities as the designs develop; and provide detailed design information based on construction plans at least 180 days prior to the start of construction, and on "as-built" designs at least 180 days before the first receipt of nuclear material. Ultimately, the design information will be captured in an IAEA Design Information

  14. Computational prediction of dust production in graphite moderated pebble bed reactors

    NASA Astrophysics Data System (ADS)

    Rostamian, Maziar

    The scope of the work reported here, which is the computational study of graphite wear behavior, supports the Nuclear Engineering University Programs project "Experimental Study and Computational Simulations of Key Pebble Bed Thermomechanics Issues for Design and Safety" funded by the US Department of Energy. In this work, modeling and simulating the contact mechanics, as anticipated in a PBR configuration, is carried out for the purpose of assessing the amount of dust generated during a full power operation year of a PBR. A methodology that encompasses finite element analysis (FEA) and micromechanics of wear is developed to address the issue of dust production and its quantification. Particularly, the phenomenon of wear and change of its rate with sliding length is the main focus of this dissertation. This work studies the wear properties of graphite by simulating pebble motion and interactions of a specific type of nuclear grade graphite, IG-11. This study consists of two perspectives: macroscale stress analysis and microscale analysis of wear mechanisms. The first is a set of FEA simulations considering pebble-pebble frictional contact. In these simulations, the mass of generated graphite particulates due to frictional contact is calculated by incorporating FEA results into Archard's equation, which is a linear correlation between wear mass and wear length. However, the experimental data by Johnson, University of Idaho, revealed that the wear rate of graphite decreases with sliding length. This is because the surfaces of the graphite pebbles become smoother over time, which results in a gradual decrease in wear rate. In order to address the change in wear rate, a more detailed analysis of wear mechanisms at room temperature is presented. In this microscale study, the wear behavior of graphite at the asperity level is studied by simulating the contact between asperities of facing surfaces. By introducing the effect of asperity removal on wear rate, a nonlinear

  15. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    NASA Astrophysics Data System (ADS)

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling

  16. KUGEL: a thermal, hydraulic, fuel performance, and gaseous fission product release code for pebble bed reactor core analysis

    SciTech Connect

    Shamasundar, B.I.; Fehrenbach, M.E.

    1981-05-01

    The KUGEL computer code is designed to perform thermal/hydraulic analysis and coated-fuel particle performance calculations for axisymmetric pebble bed reactor (PBR) cores. This computer code was developed as part of a Department of Energy (DOE)-funded study designed to verify the published core performance data on PBRs. The KUGEL code is designed to interface directly with the 2DB code, a two-dimensional neutron diffusion code, to obtain distributions of thermal power, fission rate, fuel burnup, and fast neutron fluence, which are needed for thermal/hydraulic and fuel performance calculations. The code is variably dimensioned so that problem size can be easily varied. An interpolation routine allows variable mesh size to be used between the 2DB output and the two-dimensional thermal/hydraulic calculations.

  17. PEBBLES Simulation of Static Friction and New Static Friction Benchmark

    SciTech Connect

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2010-05-01

    Pebble bed reactors contain large numbers of spherical fuel elements arranged randomly. Determining the motion and location of these fuel elements is required for calculating certain parameters of pebble bed reactor operation. This paper documents the PEBBLES static friction model. This model uses a three dimensional differential static friction approximation extended from the two dimensional Cundall and Strack model. The derivation of determining the rotational transformation of pebble to pebble static friction force is provided. A new implementation for a differential rotation method for pebble to container static friction force has been created. Previous published methods are insufficient for pebble bed reactor geometries. A new analytical static friction benchmark is documented that can be used to verify key static friction simulation parameters. This benchmark is based on determining the exact pebble to pebble and pebble to container static friction coefficients required to maintain a stable five sphere pyramid.

  18. One-dimensional modeling of radial heat removal during depressurized heatup transients in modular pebble-bed and prismatic high temperature gas-cooled reactors

    SciTech Connect

    Savage, M.G.

    1984-07-01

    A one-dimensional computational model was developed to evaluate the heat removal capabilities of both prismatic-core and pebble-bed modular HTGRs during depressurized heatup transients. A correlation was incorporated to calculate the temperature- and neutron-fluence-dependent thermal conductivity of graphite. The modified Zehner-Schluender model was used to determine the effective thermal conductivity of a pebble bed, accounting for both conduction and radiation. Studies were performed for prismatic-core and pebble-bed modular HTGRs, and the results were compared to analyses performed by GA and GR, respectively. For the particular modular reactor design studied, the prismatic HTGR peak temperature was 2152.2/sup 0/C at 38 hours following the transient initiation, and the pebble-bed peak temperature was 1647.8/sup 0/C at 26 hours. These results compared favorably with those of GA and GE, with only slight differences caused by neglecting axial heat transfer in a one-dimensional radial model. This study found that the magnitude of the initial power density had a greater effect on the temperature excursion than did the initial temperature.

  19. Aerobic degradation of a mixture of azo dyes in a packed bed reactor having bacteria-coated laterite pebbles.

    PubMed

    Senan, Resmi C; Shaffiqu, T S; Roy, J Jegan; Abraham, T Emilia

    2003-01-01

    A microbial consortium capable of aerobic degradation of a mixture of azo dyes consisting of two isolated strains (RRL,TVM) and one known strain of Pseudomonas putida (MTCC 1194) was immobilized on laterite stones. The amount of bacterial biomass attached to the laterite stones was 8.64 g per 100 g of the stone on a dry weight basis. The packed bed reactor was filled with these stones and had a total capacity of 850 mL and a void volume of 210 mL. The feed consisted of an equal mixture of seven azo dyes both in water as well as in a simulated textile effluent, at a pH of 9.0 and a salinity of 900 mg/L. The dye concentrations of influent were 25, 50, and 100 microg/mL. The residence time was varied between 0.78 and 6.23 h. It was found that at the lowest residence time 23.55, 45.73, and 79.95 microg of dye was degraded per hour at an initial dye concentration of 25, 50, and 100 microg, respectively. The pH was reduced from 9.0 to 7.0. Simulated textile effluent containing 50 microg/mL dye was degraded by 61.7%. Analysis of degradation products by TLC and HPLC showed that the dye mixture was degraded to nontoxic smaller molecules. The bacteria-coated pebbles were stable, there was no washout even after 2 months, and the reactor was found to be suitable for the aerobic degradation of azo dyes. PMID:12675610

  20. Aerobic degradation of a mixture of azo dyes in a packed bed reactor having bacteria-coated laterite pebbles.

    PubMed

    Senan, Resmi C; Shaffiqu, T S; Roy, J Jegan; Abraham, T Emilia

    2003-01-01

    A microbial consortium capable of aerobic degradation of a mixture of azo dyes consisting of two isolated strains (RRL,TVM) and one known strain of Pseudomonas putida (MTCC 1194) was immobilized on laterite stones. The amount of bacterial biomass attached to the laterite stones was 8.64 g per 100 g of the stone on a dry weight basis. The packed bed reactor was filled with these stones and had a total capacity of 850 mL and a void volume of 210 mL. The feed consisted of an equal mixture of seven azo dyes both in water as well as in a simulated textile effluent, at a pH of 9.0 and a salinity of 900 mg/L. The dye concentrations of influent were 25, 50, and 100 microg/mL. The residence time was varied between 0.78 and 6.23 h. It was found that at the lowest residence time 23.55, 45.73, and 79.95 microg of dye was degraded per hour at an initial dye concentration of 25, 50, and 100 microg, respectively. The pH was reduced from 9.0 to 7.0. Simulated textile effluent containing 50 microg/mL dye was degraded by 61.7%. Analysis of degradation products by TLC and HPLC showed that the dye mixture was degraded to nontoxic smaller molecules. The bacteria-coated pebbles were stable, there was no washout even after 2 months, and the reactor was found to be suitable for the aerobic degradation of azo dyes.

  1. Deleterious Thermal Effects due to Randomized Flow Paths in Pebble Bed, and Particle Bed Style Reactors

    NASA Technical Reports Server (NTRS)

    Moran, Robert P.

    2013-01-01

    Reactor fuel rod surface area that is perpendicular to coolant flow direction (+S) i.e. perpendicular to the P creates areas of coolant stagnation leading to increased coolant temperatures resulting in localized changes in fluid properties. Changes in coolant fluid properties caused by minor increases in temperature lead to localized reductions in coolant mass flow rates leading to localized thermal instabilities. Reductions in coolant mass flow rates result in further increases in local temperatures exacerbating changes to coolant fluid properties leading to localized thermal runaway. Unchecked localized thermal runaway leads to localized fuel melting. Reactor designs with randomized flow paths are vulnerable to localized thermal instabilities, localized thermal runaway, and localized fuel melting.

  2. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    SciTech Connect

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2009-09-01

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  3. Development Status of the PEBBLES Code for Pebble Mechanics: Improved Physical Models and Speed-up

    SciTech Connect

    Joshua J. Cogliati; Abderrafi M. Ougouag

    2009-12-01

    PEBBLES is a code for simulating the motion of all the pebbles in a pebble bed reactor. Since pebble bed reactors are packed randomly and not precisely placed, the location of the fuel elements in the reactor is not deterministically known. Instead, when determining operating parameters the motion of the pebbles can be simulated and stochastic locations can be found. The PEBBLES code can output information relevant for other simulations of the pebble bed reactors such as the positions of the pebbles in the reactor, packing fraction change in an earthquake, and velocity profiles created by recirculation. The goal for this level three milestone was to speedup the PEBBLES code through implementation on massively parallel computer. Work on this goal has resulted in speeding up both the single processor version and creation of a new parallel version of PEBBLES. Both the single processor version and the parallel running capability of the PEBBLES code have improved since the fiscal year start. The hybrid MPI/OpenMP PEBBLES version was created this year to run on the increasingly common cluster hardware profile that combines nodes with multiple processors that share memory and a cluster of nodes that are networked together. The OpenMP portions use the Open Multi-Processing shared memory parallel processing model to split the task across processors in a single node that shares memory. The Message Passing Interface (MPI) portion uses messages to communicate between different nodes over a network. The following are wall clock speed up for simulating an NGNP-600 sized reactor. The single processor version runs 1.5 times faster compared to the single processor version at the beginning of the fiscal year. This speedup is primarily due to the improved static friction model described in the report. When running on 64 processors, the new MPI/OpenMP hybrid version has a wall clock speed up of 22 times compared to the current single processor version. When using 88 processors, a

  4. Pebble Bed Reactor Power Systems for Lunar Outposts: Long Operation Life and End-of Life Storage

    NASA Astrophysics Data System (ADS)

    El-Genk, Mohamed S.; Schriener, Timothy M.

    2010-09-01

    The Pellet Bed Reactor(PeBR) and power system for supporting future lunar outposts offer many desirable design, operation and safety features and address post operation storage of spent nuclear fuel. In addition to its long, full power operation life of 66 year, the PeBR is launched without fuel and loaded after placement below grade on the lunar surface with spherical fuel pellets, designed to fully contain fission products. The fuel pellets(~1.0 cm dia.) are launched separately in subcritical canisters. The post-operation PeBR is stored below grade for > 300 year to allow the radioactivity in the spent fuel to decay to a sufficiently low level. The PeBR power system, designed for avoidance of single point failures in reactor cooling and energy conversion, nominally generates ~100 kWe at a thermal efficiency of ~ 21%. In addition to the sectored reactor core, it uses three Closed Brayton Cycle loops with centrifugal flow turbo-machines for energy conversion and He-Xe(40 g/mol) binary gas mixture working fluid and reactor coolant.

  5. Thermal decomposition of expanded polystyrene in a pebble bed reactor to get higher liquid fraction yield at low temperatures

    SciTech Connect

    Chauhan, R.S. Gopinath, S.; Razdan, P.; Delattre, C.; Nirmala, G.S.; Natarajan, R.

    2008-11-15

    Expanded polystyrene is one of the polymers produced in large quantities due to its versatile application in different fields. This polymer is one of the most intractable components in municipal solid waste. Disposal of polymeric material by pyrolysis or catalytic cracking yields valuable hydrocarbon fuels or monomers. Literature reports different types of reactors and arrangements that have uniform temperatures during pyrolysis and catalytic cracking. The present study focuses on reducing the temperature to maximize the quantity of styrene monomer in the liquid product. A bench scale reactor has been developed to recover the styrene monomer and other valuable chemicals. Experiments were carried under partial oxidation and vacuum conditions in the temperature range of 300-500 deg. C. In the pyrolysis optimization studies, the best atmospheric condition was determined to be vacuum, the pyrolysis temperature should be 500 deg. C, yield of liquid product obtained was 91.7% and yield of styrene obtained was 85.5%. In the characterization studies, distillation and IR spectroscopy experiments were carried out. The remaining of the liquid product comprises of benzene, ethyl benzene, and styrene dimers and trimers.

  6. Image reconstruction of single photon emission computed tomography (SPECT) on a pebble bed reactor (PBR) using expectation maximization and exact inversion algorithms: Comparison study by means of numerical phantom

    NASA Astrophysics Data System (ADS)

    Razali, Azhani Mohd; Abdullah, Jaafar

    2015-04-01

    Single Photon Emission Computed Tomography (SPECT) is a well-known imaging technique used in medical application, and it is part of medical imaging modalities that made the diagnosis and treatment of disease possible. However, SPECT technique is not only limited to the medical sector. Many works are carried out to adapt the same concept by using high-energy photon emission to diagnose process malfunctions in critical industrial systems such as in chemical reaction engineering research laboratories, as well as in oil and gas, petrochemical and petrochemical refining industries. Motivated by vast applications of SPECT technique, this work attempts to study the application of SPECT on a Pebble Bed Reactor (PBR) using numerical phantom of pebbles inside the PBR core. From the cross-sectional images obtained from SPECT, the behavior of pebbles inside the core can be analyzed for further improvement of the PBR design. As the quality of the reconstructed image is largely dependent on the algorithm used, this work aims to compare two image reconstruction algorithms for SPECT, namely the Expectation Maximization Algorithm and the Exact Inversion Formula. The results obtained from the Exact Inversion Formula showed better image contrast and sharpness, and shorter computational time compared to the Expectation Maximization Algorithm.

  7. Image reconstruction of single photon emission computed tomography (SPECT) on a pebble bed reactor (PBR) using expectation maximization and exact inversion algorithms: Comparison study by means of numerical phantom

    SciTech Connect

    Razali, Azhani Mohd Abdullah, Jaafar

    2015-04-29

    Single Photon Emission Computed Tomography (SPECT) is a well-known imaging technique used in medical application, and it is part of medical imaging modalities that made the diagnosis and treatment of disease possible. However, SPECT technique is not only limited to the medical sector. Many works are carried out to adapt the same concept by using high-energy photon emission to diagnose process malfunctions in critical industrial systems such as in chemical reaction engineering research laboratories, as well as in oil and gas, petrochemical and petrochemical refining industries. Motivated by vast applications of SPECT technique, this work attempts to study the application of SPECT on a Pebble Bed Reactor (PBR) using numerical phantom of pebbles inside the PBR core. From the cross-sectional images obtained from SPECT, the behavior of pebbles inside the core can be analyzed for further improvement of the PBR design. As the quality of the reconstructed image is largely dependent on the algorithm used, this work aims to compare two image reconstruction algorithms for SPECT, namely the Expectation Maximization Algorithm and the Exact Inversion Formula. The results obtained from the Exact Inversion Formula showed better image contrast and sharpness, and shorter computational time compared to the Expectation Maximization Algorithm.

  8. Postirradiation examination of beryllium pebbles

    SciTech Connect

    Gelles, D.S.

    1998-03-01

    Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements.

  9. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    DOEpatents

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  10. PEBBLES Operation and Theory Manual

    SciTech Connect

    Joshua J. Cogliati

    2010-09-01

    The PEBBLES manual describes the PEBBLES code. The PEBBLES code is a computer program designed to simulation the motion, packing and vibration of spheres that undergo various mechanical forces including gravitation, Hooke’s law force and various friction forces. The frictional forces include true static friction that allows non-zero angles of repose. Each pebble is individually simulated using the distinct element method.

  11. PEBBLES Operation and Theory Manual

    SciTech Connect

    Joshua J. Cogliati

    2011-02-01

    The PEBBLES manual describes the PEBBLES code. The PEBBLES code is a computer program designed to simulation the motion, packing and vibration of spheres that undergo various mechanical forces including gravitation, Hooke’s law force and various friction forces. The frictional forces include true static friction that allows non-zero angles of repose. Each pebble is individually simulated using the distinct element method.

  12. OPERATION PEBBLE. SUMMARY REPORT.

    ERIC Educational Resources Information Center

    NORRED, ROBERT B.

    A COMPREHENSIVE EVALUATION OF OPERATION PEBBLE'S 3 YEAR SUMMER PROGRAM FOR ECONOMICALLY AND EDUCATIONALLY DEPRIVED CHILDREN OF THE UPPER CUMBERLAND REGION OF TENNESSEE IS PRESENTED. THE INTENT OF THE PROJECT WAS TO INVOLVE THE CHILDREN IN EXPERIENCES THAT MIGHT EXPAND THE HORIZONS OF THEIR STAGNANT, HIGHLY STRUCTURED CULTURAL ENVIRONMENT, WITHOUT…

  13. Pebble Puzzle Solved

    NASA Technical Reports Server (NTRS)

    2004-01-01

    [figure removed for brevity, see original site] Figure 1 In the quest to determine if a pebble was jamming the rock abrasion tool on NASA's Mars Exploration Rover Opportunity, scientists and engineers examined this up-close, approximate true-color image of the tool. The picture was taken by the rover's panoramic camera, using filters centered at 601, 535, and 482 nanometers, at 12:47 local solar time on sol 200 (August 16, 2004).

    Colored spots have been drawn on this image corresponding to regions where panoramic camera reflectance spectra were acquired (see chart in Figure 1). Those regions are: the grinding wheel heads (yellow); the rock abrasion tool magnets (green); the supposed pebble (red); a sunlit portion of the aluminum rock abrasion tool housing (purple); and a shadowed portion of the rock abrasion tool housing (brown). These spectra demonstrated that the composition of the supposed pebble was clearly different from that of the sunlit and shadowed portions of the rock abrasion tool, while similar to that of the dust-coated rock abrasion tool magnets and grinding heads. This led the team to conclude that the object disabling the rock abrasion tool was indeed a martian pebble.

  14. Reprocessing of lithium titanate pebbles by graphite bed method

    NASA Astrophysics Data System (ADS)

    Hong, Ming; Zhang, Yingchun; Xiang, Maoqiao; Zhang, Yun

    2015-04-01

    Lithium titanate enriched by 6Li isotope is considered as a candidate of tritium breeding materials for fusion reactors due to its excellent performance. The reuse of burned Li2TiO3 pebbles is an important issue because of the high costs of 6Li-enriched materials and waste considerations. For this purpose, reprocessing of Li2TiO3 pebbles by graphite bed method was developed. Simulative Li2TiO3 pebbles with low-lithium content according to the expected lithium burn-up were fabricated. After that, Li2TiO3 pebbles were re-fabricated with lithium carbonate as lithium additives, in order to gain the composition of lithium titanate with a Li/Ti ratio of 2. The process was optimized to obtain reprocessed Li2TiO3 pebbles that were suitable for reuse as ceramic breeder. Density, porosity, grain size and crushing load of the reprocessed pebbles were characterized. This process did not deteriorate the properties of the reprocessed pebbles and was almost no waste generation.

  15. From Pebbles to Planets

    NASA Astrophysics Data System (ADS)

    Johansen, Anders

    2013-10-01

    Planets form in protoplanetary discs around young stars as dust and ice particles collide to form ever larger bodies. Particle concentration in the turbulent gas flow may be necessary to form the planetesimals which are the building blocks of both the terrestrial planets and the cores of the gas giants and the ice giants. The streaming instability, which feeds off the relative motion of gas and particles, is a powerful mechanism to create overdense particle filaments. These filaments contract under their own gravity to form planetesimals with a wide range of sizes. I will also discuss how the pebbles left over from the planetesimal formation stage can lead to rapid formation of the cores of gas giants, well within the protoplanetary disc life-time, even in wide orbits.

  16. Comet Formation in Collapsing Pebble Clouds: Pebble Formation

    NASA Astrophysics Data System (ADS)

    Lorek, Sebastian; Lacerda, Pedro; Blum, Jürgen

    2016-10-01

    The formation of comets by gradual growth from (sub-)micron sized ice and dust monomers to km-sized bodies suffers from growth barriers (bouncing, fragmentation, drift). Growth stalls at sizes between mm and m, rendering it considerably difficult to form km-sized objects. However, the streaming instability and subsequent gravitational collapse of clouds of pebbles (particle agglomerates) provide an alternative. The pebbles require Stokes numbers between 0.01 and 3, which corresponds to sizes between mm and dm, unless the pebbles are very porous. Furthermore, the local solid/gas density ratio must be near unity and the local total mass in solids must be >2-3x higher than the minimum mass solar nebula value (1% of gas mass). The gravitational collapse of the pebble clouds then bypasses the growth barriers, forming km-sized bodies directly. The observed bulk properties of comets, e.g. porosity near 80%, are consistent with this scenario. Okuzumi et al. (2012) showed that including porosity comets can form directly via coagulation from sub-micron monomers. However, this relies on using 0.1 micron monomers and pure sticking collisions. Krijt et al. (2015) included erosion and found that highly porous pebbles around 109 g in mass can form and might trigger the streaming instability. Drazkowska & Dullemond (2014) showed that compact coagulation can lead to triggering the streaming instability. All those studies include only ice and a simplified collision model. However, a large fraction of a comet's mass is dust. Here, we develop a pebble formation model that includes sticking, bouncing, mass transfer/erosion, and fragmentation, as well as porosity. To take dust and ice into account, we extended the collision model for the treatment of mixed pebbles by linearly interpolating the threshold velocities and compression curves between the cases of pure dust and pure ice based on the fractional abundance of dust monomers. Our simulations show that pebble formation with the full

  17. TEM study of impurity segregations in beryllium pebbles

    NASA Astrophysics Data System (ADS)

    Klimenkov, M.; Chakin, V.; Moeslang, A.; Rolli, R.

    2014-12-01

    Beryllium is planned to be used as a neutron multiplier in the Helium-cooled Pebble Bed European concept of a breeding blanket of demonstration power reactor DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron-irradiated at temperatures typical of fusion blankets. Beryllium pebbles 1 mm in diameter produced by the rotating electrode method were subjected to a TEM study before and after irradiation at High Flux Reactor, Petten, Netherlands at 861 K. The grain size varied in a wide range from sub-micron size up to several tens of micrometers, which indicated formation bimodal grain size distribution. Based on the application of combined electron energy loss spectroscopy and energy dispersive X-ray spectroscopy methods, we suggest that impurity precipitates play an important role in controlling the mechanical properties of beryllium. The impurity elements were present in beryllium at a sub-percent concentration form beryllide particles of a complex (Fe/Al/Mn/Cr)B composition. These particles are often ordered along dislocations lines, forming several micron-long chains. It can be suggested that fracture surfaces often extended along these chains in irradiated material.

  18. The shapes of beach pebbles

    USGS Publications Warehouse

    Wentworth, Chester K.

    1923-01-01

    There is much confusion in geologic literature as to the shapes of fluvial and beach pebbles and the differences between them, if differences exist. Though the contrary has been asserted, most geologists who have written on the subject appear to hold the view that beach pebbles are generally flatter than river pebbles, having discoid, lozenge-shaped, ellipsoid, or oval forms. It is asserted by some that these forms are produced by pushing of the rock fragments to and fro by the waves. Others have considered that the shapes of the original fragments and the inherent structure of the rock are dominant in determining the shapes of beach pebbles, and with this view the writer is in accord. That beach pebbles, even those composed of massive igneous rocks are commonly of a flattened oval form seems certain, as has been stated elsewhere, but this fact is probably to be attributed to the development of such forms from original flat fragments or from rocks of schistose structure or to the segregation of such forms under the peculiar action of the waves, rather than to their production by a specialized wave abrasion.

  19. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

    This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

  20. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    SciTech Connect

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  1. HTR-PROTEUS Pebble Bed Experimental Program Cores 1, 1A, 2, and 3: Hexagonal Close Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    SciTech Connect

    John D. Bess; Barbara H. Dolphin; James W. Sterbentz; Luka Snoj; Igor Lengar; Oliver Köberl

    2012-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. Four benchmark experiments were evaluated in this report: Cores 1, 1A, 2, and 3. These core configurations represent the hexagonal close packing (HCP) configurations of the HTR-PROTEUS experiment with a moderator-to-fuel pebble ratio of 1:2. Core 1 represents the only configuration utilizing ZEBRA control rods. Cores 1A, 2, and 3 use withdrawable, hollow, stainless steel control rods. Cores 1 and 1A are similar except for the use of different control rods; Core 1A also has one less layer of pebbles (21 layers instead of 22). Core 2 retains the first 16 layers of pebbles from Cores 1 and 1A and has 16 layers of moderator pebbles stacked above the fueled layers. Core 3 retains the first 17 layers of pebbles but has polyethylene rods inserted between pebbles to simulate water ingress. The additional partial pebble layer (layer 18) for Core 3 was not included as it was used for core operations and not the reported critical configuration. Cores 1, 1A, 2, and 3 were determined to be acceptable benchmark experiments.

  2. Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

    DOE PAGES

    Bess, John D.; Montierth, Leland; Köberl, Oliver; Snoj, Luka

    2014-10-09

    Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the ²³⁵U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of keff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greatermore » than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of keff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  3. Benchmark Evaluation of HTR-PROTEUS Pebble Bed Experimental Program

    SciTech Connect

    Bess, John D.; Montierth, Leland; Köberl, Oliver; Snoj, Luka

    2014-10-09

    Benchmark models were developed to evaluate 11 critical core configurations of the HTR-PROTEUS pebble bed experimental program. Various additional reactor physics measurements were performed as part of this program; currently only a total of 37 absorber rod worth measurements have been evaluated as acceptable benchmark experiments for Cores 4, 9, and 10. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the ²³⁵U enrichment of the fuel, impurities in the moderator pebbles, and the density and impurity content of the radial reflector. Calculations of keff with MCNP5 and ENDF/B-VII.0 neutron nuclear data are greater than the benchmark values but within 1% and also within the 3σ uncertainty, except for Core 4, which is the only randomly packed pebble configuration. Repeated calculations of keff with MCNP6.1 and ENDF/B-VII.1 are lower than the benchmark values and within 1% (~3σ) except for Cores 5 and 9, which calculate lower than the benchmark eigenvalues within 4σ. The primary difference between the two nuclear data libraries is the adjustment of the absorption cross section of graphite. Simulations of the absorber rod worth measurements are within 3σ of the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  4. Fractured pebbles - A new stress indicator

    SciTech Connect

    Eidelman, A.; Reches, Z. )

    1992-04-01

    Well-organized patterns of tensile fractures were found in pebbles of young conglomerates in the Salton trough, California, and in the Dead Sea rift, Israel. The fractures are subparallel to each other in a single pebble and are within numerous pebbles in an outcrop. The authors show that intrapebble tension could develop due to the amplification of the stresses inside a competent pebble embedded within a compliant matrix; furthermore, tensile stresses may form in the pebbles, even under compressive tectonic stresses. The regional trends of the fractures are consistent in each of the study areas, and they appear to be excellent indicators of the tectonic stresses. The derived angles between {sigma}{sub hmax} (the largest horizontal compression) and the dominant local strike-slip fault is about 40{degree} in the En Yahav region of the Dead Sea rift, and about 75{degree} in the Indio Hills area of the Salton trough; these angles are in agreement with other stress data.

  5. Experimental investigation of the pebble bed structure by using gamma ray tomography

    NASA Astrophysics Data System (ADS)

    Ahmed, Fadha Shakir

    Pebble Bed Reactors offer a future for new nuclear energy plants. They are small, inherently safe, and can be competitive with fossil fuels. The fuel forms a randomly stacked pebble with non-uniform fuel densities. The thermal-mechanical behavior of pebble bed reactor core is depends strongly on the spatial variation of packing fraction in the bed and in particular on the number of contacts between pebbles, and between the pebbles and the blanket walls. To investigate these effects, experimental data to characterize bed structure are needed along with other numerical simulation and computational tools for validation. In this study, a powerful technique of high-energy gamma-ray computed tomography (CT scanner system) is employed for the first time for the quantification of the structure of pebble bed in term of the cross-sectional time-averaged void and distributions, it radial profiles and the statistical analysis. The alternative minimization (AM) iteration algorithm is used for image reconstruction. The spatial resolution of the CT scan is about 2 mm with 100 x 100 pixel used to reconstruct the cross-sectional image. Results of tomography with this advanced technique on three different pebble sizes at different axial levels are presented. The bed consisted of a glass spheres (Marbles) with a diameter d1= 1.27 cm, d2= 2.54 cm and d3= 5 cm in a Plexiglas cylinder with diameter D = 30.48 cm (D/d1 = 24, D/d2 = 12 and D/d3 = 6), and had an average void fraction epsilon1= 0.389, epsilon2 = 0.40 and epsilon 3 =0.43, respectively. The radial void fraction profile showed large oscillations with the bigger pebble diameters and the void fraction is higher on the wall with a minimum void fraction of 0.33 at 0.68 pebble diameter away from the wall. It was found that the void distribution in random packed bed depends strongly on the pebble diameter with respect to the bed diameter (D/d p) and the packing mode. The oscillation is quiet large with the smaller aspect ratio (D

  6. Formation of planetesimals in collapsing pebble clouds

    NASA Astrophysics Data System (ADS)

    Wahlberg Jansson, K.; Johansen, A.

    2014-07-01

    Asteroids and Kuiper belt objects are remnant planetesimals from the epoch of planet formation. Their physical properties hold important clues to understanding how minor bodies formed in the Solar Nebula. The first stage of the planet formation process is the accumulation of dust and ice grains into mm-cm-sized pebbles. Due to the interaction with the gas in the protoplanetary disk, these pebbles can clump together through the streaming instability and form gravitationally bound particle pebble 'clouds'. Pebbles in the cloud collide with each other, dissipating energy into heat. As the cloud loses energy, it contracts, and one would expect the particles to move faster and faster due to the negative heat capacity nature of self-gravitating systems. However, for high-mass clouds, the collapse is limited by free-fall and the cloud does not have time to virialize. This in turn leads to lower collision speeds but thanks to increased density also to increased collision rates and a runaway collapse. We investigate three important properties of the collapse: (i) the time-scale to collapse to solid density, (ii) the temporal evolution of the size spectrum of the pebbles, and (iii) the multiplicity of the resulting planetesimals. We find that planetesimals larger than 100 km in radius collapse on the free-fall time-scale of about 25 years. Lower-mass clouds have longer pebble collision time-scales and hence collapse much more slowly, with collapse times of a few hundred years for 10-km-scale planetesimals and a few thousand years for 1-km-scale planetesimals. The mass of the pebble cloud also determines the structure of the resulting planetesimal. The collision speed among the pebbles in low- mass clouds is below the threshold for fragmentation, forming pebble- pile planetesimals consisting of the primordial pebbles from the nebula. Planetesimals above 100 km in radius, on the other hand, consist of mixtures of dust (pebble fragments) and pebbles which have undergone

  7. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess; Leland M. Montierth

    2014-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  8. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORE 4: RANDOM PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess; Leland M. Montierth

    2013-03-01

    In its deployment as a pebble bed reactor (PBR) critical facility from 1992 to 1996, the PROTEUS facility was designated as HTR-PROTEUS. This experimental program was performed as part of an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on the Validation of Safety Related Physics Calculations for Low Enriched HTGRs. Within this project, critical experiments were conducted for graphite moderated LEU systems to determine core reactivity, flux and power profiles, reaction-rate ratios, the worth of control rods, both in-core and reflector based, the worth of burnable poisons, kinetic parameters, and the effects of moisture ingress on these parameters. One benchmark experiment was evaluated in this report: Core 4. Core 4 represents the only configuration with random pebble packing in the HTR-PROTEUS series of experiments, and has a moderator-to-fuel pebble ratio of 1:1. Three random configurations were performed. The initial configuration, Core 4.1, was rejected because the method for pebble loading, separate delivery tubes for the moderator and fuel pebbles, may not have been completely random; this core loading was rejected by the experimenters. Cores 4.2 and 4.3 were loaded using a single delivery tube, eliminating the possibility for systematic ordering effects. The second and third cores differed slightly in the quantity of pebbles loaded (40 each of moderator and fuel pebbles), stacked height of the pebbles in the core cavity (0.02 m), withdrawn distance of the stainless steel control rods (20 mm), and withdrawn distance of the autorod (30 mm). The 34 coolant channels in the upper axial reflector and the 33 coolant channels in the lower axial reflector were open. Additionally, the axial graphite fillers used in all other HTR-PROTEUS configurations to create a 12-sided core cavity were not used in the randomly packed cores. Instead, graphite fillers were placed on the cavity floor, creating a funnel-like base, to discourage ordering

  9. Molecular dynamics simulation for PBR pebble tracking simulation via a random walk approach using Monte Carlo simulation.

    PubMed

    Lee, Kyoung O; Holmes, Thomas W; Calderon, Adan F; Gardner, Robin P

    2012-05-01

    Using a Monte Carlo (MC) simulation, random walks were used for pebble tracking in a two-dimensional geometry in the presence of a biased gravity field. We investigated the effect of viscosity damping in the presence of random Gaussian fluctuations. The particle tracks were generated by Molecular Dynamics (MD) simulation for a Pebble Bed Reactor. The MD simulations were conducted in the interaction of noncohesive Hertz-Mindlin theory where the random walk MC simulation has a correlation with the MD simulation. This treatment can easily be extended to include the generation of transient gamma-ray spectra from a single pebble that contains a radioactive tracer. Then the inverse analysis thereof could be made to determine the uncertainty of the realistic measurement of transient positions of that pebble by any given radiation detection system designed for that purpose.

  10. Fabrication and characterization of LiH ceramic pebbles by wet process

    NASA Astrophysics Data System (ADS)

    Xiang, Maoqiao; Zhang, Yingchun; Hong, Ming; Liu, Zhiang; Leng, Jiaxun; Zhang, Yun; Zhang, Jialiang; Wang, Wenchang

    2014-09-01

    Lithium hydride (LiH) ceramic pebbles, a new potential tritium breeding material in fusion-fission or fusion reactor blanket, were prepared by wet process for the first time. XRD results showed that LiOH, LiOH·H2O, Li2CO3 and Li2O were found in the surface of LiH pebbles. However, the pure phase of LiH pebbles without cracks could be obtained by paraffin wax coating technique. The average value (a.v.) of the sphericity and the diameter were 1.01 and 0.98 mm, respectively. The LiH pebbles sintered at 450 °C for 3 h under 80 ml/min flowing argon, reached ∼92.3% of the theoretical density, with the grain size of 5.59 μm (a.v.). And the crush load was measured to be 15 N on average. The described wet process exhibited multiple advantages for fabricating LiH pebbles.

  11. "Smart pebble" designs for sediment transport monitoring

    NASA Astrophysics Data System (ADS)

    Valyrakis, Manousos; Alexakis, Athanasios; Pavlovskis, Edgars

    2015-04-01

    Sediment transport, due to primarily the action of water, wind and ice, is one of the most significant geomorphic processes responsible for shaping Earth's surface. It involves entrainment of sediment grains in rivers and estuaries due to the violently fluctuating hydrodynamic forces near the bed. Here an instrumented particle, namely a "smart pebble", is developed to investigate the exact flow conditions under which individual grains may be entrained from the surface of a gravel bed. This could lead in developing a better understanding of the processes involved, focusing on the response of the particle during a variety of flow entrainment events. The "smart pebble" is a particle instrumented with MEMS sensors appropriate for capturing the hydrodynamic forces a coarse particle might experience during its entrainment from the river bed. A 3-axial gyroscope and accelerometer registers data to a memory card via a microcontroller, embedded in a 3D-printed waterproof hollow spherical particle. The instrumented board is appropriately fit and centred into the shell of the pebble, so as to achieve a nearly uniform distribution of the mass which could otherwise bias its motion. The "smart pebble" is powered by an independent power to ensure autonomy and sufficiently long periods of operation appropriate for deployment in the field. Post-processing and analysis of the acquired data is currently performed offline, using scientific programming software. The performance of the instrumented particle is validated, conducting a series of calibration experiments under well-controlled laboratory conditions.

  12. "Smart pebble" design for environmental monitoring applications

    NASA Astrophysics Data System (ADS)

    Valyrakis, Manousos; Pavlovskis, Edgars

    2014-05-01

    Sediment transport, due to primarily the action of water, wind and ice, is one of the most significant geomorphic processes responsible for shaping Earth's surface. It involves entrainment of sediment grains in rivers and estuaries due to the violently fluctuating hydrodynamic forces near the bed. Here an instrumented particle, namely a "smart pebble", is developed to investigate the exact flow conditions under which individual grains may be entrained from the surface of a gravel bed. This could lead in developing a better understanding of the processes involved, while focusing on the response of the particle during a variety of flow entrainment events. The "smart pebble" is a particle instrumented with MEMS sensors appropriate for capturing the hydrodynamic forces a coarse particle might experience during its entrainment from the river bed. A 3-axial gyroscope and accelerometer registers data to a memory card via a microcontroller, embedded in a 3D-printed waterproof hollow spherical particle. The instrumented board is appropriately fit and centred into the shell of the pebble, so as to achieve a nearly uniform distribution of the mass which could otherwise bias its motion. The "smart pebble" is powered by an independent power to ensure autonomy and sufficiently long periods of operation appropriate for deployment in the field. Post-processing and analysis of the acquired data is currently performed offline, using scientific programming software. The performance of the instrumented particle is validated, conducting a series of calibration experiments under well-controlled laboratory conditions. "Smart pebble" allows for a wider range of environmental sensors (e.g. for environmental/pollutant monitoring) to be incorporated so as to extend the range of its application, enabling accurate environmental monitoring which is required to ensure infrastructure resilience and preservation of ecological health.

  13. Pebble Accretion and the Diversity of Planetary Systems

    NASA Astrophysics Data System (ADS)

    Chambers, J. E.

    2016-07-01

    This paper examines the standard model of planet formation, including pebble accretion, using numerical simulations. Planetary embryos that are large enough to become giant planets do not form beyond the ice line within a typical disk lifetime unless icy pebbles stick at higher speeds than in experiments using rocky pebbles. Systems like the solar system (small inner planets and giant outer planets) can form if icy pebbles are stickier than rocky pebbles, and if the planetesimal formation efficiency increases with pebble size, which prevents the formation of massive terrestrial planets. Growth beyond the ice line is dominated by pebble accretion. Most growth occurs early, when the surface density of the pebbles is high due to inward drift of the pebbles from the outer disk. Growth is much slower after the outer disk is depleted. The outcome is sensitive to the disk radius and turbulence level, which control the lifetime and maximum size of pebbles. The outcome is sensitive to the size of the largest planetesimals because there is a threshold mass for the onset of pebble accretion. The planetesimal formation rate is unimportant, provided that some large planetesimals form while the pebbles remain abundant. Two outcomes are seen, depending on whether pebble accretion begins while the pebbles are still abundant. Either multiple gas-giant planets form beyond the ice line, small planets form close to the star, and a Kuiper-belt-like disk of bodies is scattered outward by the giant planets; or no giants form and the bodies remain an Earth-mass or smaller.

  14. Pebbles and Cobbles at MPF Site

    NASA Technical Reports Server (NTRS)

    1997-01-01

    Pebbles are seen in lander images, along with cobbles. For example, in this picture, we see the same pebbles that were visible in the Sojourner rover image of the 'Cabbage Patch' (PIA00984). In addition, a cobble within the rock 'Lamb' (upper left) is apparent. This indicates that Lamb may be a conglomerate (Lamb is 0.32 m x 0.15 m).

    Mars Pathfinder is the second in NASA's Discovery program of low-cost spacecraft with highly focused science goals. The Jet Propulsion Laboratory, Pasadena, CA, developed and manages the Mars Pathfinder mission for NASA's Office of Space Science, Washington, D.C. JPL is a division of the California Institute of Technology (Caltech). The Imager for Mars Pathfinder (IMP) was developed by the University of Arizona Lunar and Planetary Laboratory under contract to JPL. Peter Smith is the Principal Investigator.

  15. Tritium analyses of COBRA-1A2 beryllium pebbles

    SciTech Connect

    Baldwin, D.L.

    1998-03-01

    Selected tritium measurements have been completed for the COBRA-1A2 experiment C03 and D03 beryllium pebbles. The completed results, shown in Tables 1, 2, and 3, include the tritium assay results for the 1-mm and 3-mm C03 pebbles, and the 1-mm D03 pebbles, stepped anneal test results for both types of 1-mm pebbles, and the residual analyses for the stepped-anneal specimens. All results have been reported with date-of-count and are not corrected for decay. Stepped-anneal tritium release response is provided in addenda.

  16. Pebble fuel design for the PB-FHR

    SciTech Connect

    Cisneros, A. T.; Scarlat, R. O.; Laufer, M. R.; Greenspan, E.; Peterson, P. F.

    2012-07-01

    This paper presents the results of parametric studies of pebble fuel that can guide the design of future PB-FHR cores. The pebble fuel designs are assessed using the following performance characteristics: burnup, reactivity feedback, transient response, timescale to reach equilibrium cycle, and protection of structural components. The performance of a thorium pebble blanket is assessed by comparing against a seed-only system and system that utilizes a graphite pebble reflector instead of a thorium blanket. This paper presents the functional requirements and a methodology to assess these fuel pebble designs. This paper identifies a feasible design space for low enriched uranium pebbles and selected a baseline pebble design for safe, economic energy generation. Furthermore, this study finds a thorium blanket does not increase the performance of the system significantly with respect to a graphite pebble reflector. Therefore, a graphite pebble reflector is recommended in the baseline full-core design to extend the lifetime of the outer solid graphite reflector to the life of plant. (authors)

  17. Fabrication of Li4SiO4 pebbles by wet method with modified powders synthesized via sol-gel process

    NASA Astrophysics Data System (ADS)

    Yu, Xiaohe; Yang, Mao; Lu, Tiecheng; Wei, Nian; Wei, Jianjun; Shi, Yanli; Huang, Zhangyi; Xiang, Xiaogang; Zhang, Qinghua; Zhang, Wei

    2015-01-01

    Li4SiO4 pebbles have been recognized as attractive tritium breeder materials in the fusion reactor blanket of international thermonuclear experimental reactor (ITER). In this work, we present a facile method to prepare Li4SiO4 pebbles of high density and sphericity by using a directive wet method with the Li4SiO4 powders synthesized via sol-gel process. The Li4SiO4 powders were prepared with two-step calcinating method, followed by a ball-milling process. Thermal and phase analysis, morphologies and sintering behaviors observations of the pebbles were carried out systematically. Experimental results show that the pure phase powders with white color that prepared by using two-step calcinating method is different from the powders prepared by the traditional direct calcinating method. The subsequent ball milling process proves to be effective to improve the relative density of the sintered body. When sintered at the temperature as low as 850 °C for 4 h, the favorable Li4SiO4 pebbles with uniform size (∼1 mm), good sphericity (1.02), and high density (above 90% T.D.) were fabricated by using a directive wet method. The as-fabricated pebbles hold good potential as tritium breeding materials for blankets.

  18. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  19. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  20. Recovery and recycling of lithium value from spent lithium titanate (Li2TiO3) pebbles

    NASA Astrophysics Data System (ADS)

    Mandal, D.

    2013-09-01

    In the first generation fusion reactors the fusion of deuterium (D) and tritium (T) is considered to produce energy to meet the future energy demand. Deuterium is available in nature whereas, tritium is not. Lithium-6 (Li6) isotope has the ability to produce tritium in the n, α nuclear reaction with neutrons. Thus lithium-based ceramics enriched by Li6 isotope are considered for the tritium generation for its use in future fusion reactors. Lithium titanate is one such Li-based ceramic material being considered for its some attractive properties viz., high thermal and chemical stability, high thermal conductivity, and low tritium solubility. It is reported in the literature, that the burn up of these pebbles in the fusion reactor will be limited to only 15-17 atomic percentage. At the end of life, the pebbles will contain more than 45% unused Li6 isotope. Due to the high cost of enriched Li6 and the waste disposal considerations, it is necessary to recover the unused Li from the spent lithium titanate pebbles. Till date, only the feasibilities of different processes are reported, but no process details are available. Experiments were carried out for the recovery of Li from simulated Li2TiO3 pebbles and to reuse of lithium in lithium titanate pebble fabrication. The details of the experiments and results are discussed in this paper. Simulated lithium titanate (Li2TiO3) pebbles. The objective of the study is to develop a process which can be used to recover lithium value form the spent Li2TiO3 pebbles from future fusion reactor. The Li2TiO3 pebbles used in the study were synthesized and fabricated by the solid state reaction process developed by Mandal et al. described in details somewhere else [1,2]. Spherical Li2TiO3 pebbles of size 1.0 mm were used and the properties of the Li2TiO3 pebbles used in the study are shown in Table 1. Hydrochloric acid (HCl), of 99.8% purity, purchased from Merck and Loba Chemicals, Mumbai, India. To leach lithium from Li2TiO3

  1. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  2. Li 2TiO 3 pebbles reprocessing, recovery of 6Li as Li 2CO 3

    NASA Astrophysics Data System (ADS)

    Alvani, C.; Casadio, S.; Contini, V.; Di Bartolomeo, A.; Lulewicz, J. D.; Roux, N.

    2002-12-01

    A process for obtaining Li 2CO 3 from Li 2TiO 3 powder by wet chemistry was developed. This is considered useful in view of the recovery of the 6Li isotope from lithium titanate breeder burned to its end of life in a fusion reactor. The process was optimized with respect to the chemical attack of titanate and the precipitation of carbonate from aqueous solutions to get a powder with chemical and morphological characteristics suitable for its reexploitation in the fabrication of Li 2TiO 3 pebbles. Reprocessing was also planned to adjust the 6Li concentration to the desired value and to obtain a homogeneous distribution in the powder batch. Further development concerning reprocessing of sintered Li 2TiO 3 pebbles is in progress exploiting the results obtained with lithium titanate powders.

  3. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess

    2014-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  4. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  5. On the growth of pebble-accreting planetesimals

    NASA Astrophysics Data System (ADS)

    Visser, Rico G.; Ormel, Chris W.

    2016-02-01

    Context. Pebble accretion is a newly discovered mechanism to quickly grow the cores of planets. In pebble accretion, gravity and gas drag conspire to yield large collisional cross sections for small particles in protoplanetary disks. However, before pebble accretion commences, aerodynamic deflection may act to prevent planetesimals from becoming large, because particles tend to follow gas streamlines. Aims: We derive the planetesimal radius where pebble accretion is initiated and determine the growth timescales of planetesimals by sweep-up of small particles. Methods: The equation of motion for a pebble, including gas drag and gravitational interactions, was integrated in three dimensions at distances of 1, 3, and 10 AU from the star. We obtained the collision efficiency factor as the ratio of the numerically obtained collisional cross section to the planetesimal surface area, from which we obtained the growth timescales. Integrations were conducted in the potential flow limit (steady, inviscid) and in the Stokes flow regime (steady, viscid). Results: Only particles of stopping time ts ≪ tX where tX ≈ 103 s experience aerodynamic deflection. Even in this case, the planetesimal's gravity always ensures positive collision factors. The planetesimal radius where growth proceeds slowest is ≈ 100 km (less for colder disks) corresponding to interactions shifting from the geometric to the Safronov focusing regime. For particles ts ≫ tX pebble accretion only commences after this phase and is characterized by a steep drop in growth timescales. At 1 AU, growth timescales are shorter than the disk lifetime for pebbles larger than 0.03 cm. The planetesimal radius RPA where pebble accretion commences increases with disk orbital radius. At distances beyond ~ 10 AU, sweep-up growth times are always longer than 10 Myr, while in the inner disk (≲3 AU) the viability of the sweep-up scenario is determined by the outcome of pebble-planetesimal collisions in the geometric

  6. Estimating anisotropic diffusion of neutrons near the boundary of a pebble bed random system

    SciTech Connect

    Vasques, R.

    2013-07-01

    Due to the arrangement of the pebbles in a Pebble Bed Reactor (PBR) core, if a neutron is located close to a boundary wall, its path length probability distribution function in directions of flight parallel to the wall is significantly different than in other directions. Hence, anisotropic diffusion of neutrons near the boundaries arises. We describe an analysis of neutron transport in a simplified 3-D pebble bed random system, in which we investigate the anisotropic diffusion of neutrons born near one of the system's boundary walls. While this simplified system does not model the actual physical process that takes place near the boundaries of a PBR core, the present work paves the road to a formulation that may enable more accurate diffusion simulations of such problems to be performed in the future. Monte Carlo codes have been developed for (i) deriving realizations of the 3-D random system, and (ii) performing 3-D neutron transport inside the heterogeneous model; numerical results are presented for three different choices of parameters. These numerical results are used to assess the accuracy of estimates for the mean-squared displacement of neutrons obtained with the diffusion approximations of the Atomic Mix Model and of the recently introduced [1] Non-Classical Theory with angular-dependent path length distribution. The Non-Classical Theory makes use of a Generalized Linear Boltzmann Equation in which the locations of the scattering centers in the system are correlated and the distance to collision is not exponentially distributed. We show that the results predicted using the Non-Classical Theory successfully model the anisotropic behavior of the neutrons in the random system, and more closely agree with experiment than the results predicted by the Atomic Mix Model. (authors)

  7. A method for estimating maximum static rainfall retention in pebble mulches used for soil moisture conservation

    NASA Astrophysics Data System (ADS)

    Peng, Hongtao; Lei, Tingwu; Jiang, Zhiyun; Horton, Robert

    2016-06-01

    Mulching of agricultural fields and gardens with pebbles has long been practiced to conserve soil moisture in some semi-arid regions with low precipitation. Rainfall interception by the pebble mulch itself is an important part of the computation of the water balance for the pebble mulched fields and gardens. The mean equivalent diameter (MED) was used to characterize the pebble size. The maximum static rainfall retention in pebble mulch is based on the water penetrating into the pores of pebbles, the water adhering to the outside surfaces of pebbles and the water held between pebbles of the mulch. Equations describing the water penetrating into the pores of pebbles and the water adhering to the outside surface of pebbles are constructed based on the physical properties of water and the pebble characteristics. The model for the water between pebbles of the mulch is based on the basic equation to calculate the water bridge volume and the basic coordination number model. A method to calculate the maximum static rainfall retention in the pebble mulch is presented. Laboratory rain simulation experiments were performed to test the model with measured data. Paired sample t-tests showed no significant differences between the values calculated with the method and the measured data. The model is ready for testing on field mulches.

  8. Reconstructing the transport history of pebbles on Mars.

    PubMed

    Szabó, Tímea; Domokos, Gábor; Grotzinger, John P; Jerolmack, Douglas J

    2015-01-01

    The discovery of remarkably rounded pebbles by the rover Curiosity, within an exhumed alluvial fan complex in Gale Crater, presents some of the most compelling evidence yet for sustained fluvial activity on Mars. While rounding is known to result from abrasion by inter-particle collisions, geologic interpretations of sediment shape have been qualitative. Here we show how quantitative information on the transport distance of river pebbles can be extracted from their shape alone, using a combination of theory, laboratory experiments and terrestrial field data. We determine that the Martian basalt pebbles have been carried tens of kilometres from their source, by bed-load transport on an alluvial fan. In contrast, angular clasts strewn about the surface of the Curiosity traverse are indicative of later emplacement by rock fragmentation processes. The proposed method for decoding transport history from particle shape provides a new tool for terrestrial and planetary sedimentology.

  9. Reconstructing the transport history of pebbles on Mars

    PubMed Central

    Szabó, Tímea; Domokos, Gábor; Grotzinger, John P.; Jerolmack, Douglas J.

    2015-01-01

    The discovery of remarkably rounded pebbles by the rover Curiosity, within an exhumed alluvial fan complex in Gale Crater, presents some of the most compelling evidence yet for sustained fluvial activity on Mars. While rounding is known to result from abrasion by inter-particle collisions, geologic interpretations of sediment shape have been qualitative. Here we show how quantitative information on the transport distance of river pebbles can be extracted from their shape alone, using a combination of theory, laboratory experiments and terrestrial field data. We determine that the Martian basalt pebbles have been carried tens of kilometres from their source, by bed-load transport on an alluvial fan. In contrast, angular clasts strewn about the surface of the Curiosity traverse are indicative of later emplacement by rock fragmentation processes. The proposed method for decoding transport history from particle shape provides a new tool for terrestrial and planetary sedimentology. PMID:26460507

  10. Reconstructing the transport history of pebbles on Mars.

    PubMed

    Szabó, Tímea; Domokos, Gábor; Grotzinger, John P; Jerolmack, Douglas J

    2015-01-01

    The discovery of remarkably rounded pebbles by the rover Curiosity, within an exhumed alluvial fan complex in Gale Crater, presents some of the most compelling evidence yet for sustained fluvial activity on Mars. While rounding is known to result from abrasion by inter-particle collisions, geologic interpretations of sediment shape have been qualitative. Here we show how quantitative information on the transport distance of river pebbles can be extracted from their shape alone, using a combination of theory, laboratory experiments and terrestrial field data. We determine that the Martian basalt pebbles have been carried tens of kilometres from their source, by bed-load transport on an alluvial fan. In contrast, angular clasts strewn about the surface of the Curiosity traverse are indicative of later emplacement by rock fragmentation processes. The proposed method for decoding transport history from particle shape provides a new tool for terrestrial and planetary sedimentology. PMID:26460507

  11. Effective Thermal Property Estimation of Unitary Pebble Beds Based on a CFD-DEM Coupled Method for a Fusion Blanket

    NASA Astrophysics Data System (ADS)

    Chen, Lei; Chen, Youhua; Huang, Kai; Liu, Songlin

    2015-12-01

    Lithium ceramic pebble beds have been considered in the solid blanket design for fusion reactors. To characterize the fusion solid blanket thermal performance, studies of the effective thermal properties, i.e. the effective thermal conductivity and heat transfer coefficient, of the pebble beds are necessary. In this paper, a 3D computational fluid dynamics discrete element method (CFD-DEM) coupled numerical model was proposed to simulate heat transfer and thereby estimate the effective thermal properties. The DEM was applied to produce a geometric topology of a prototypical blanket pebble bed by directly simulating the contact state of each individual particle using basic interaction laws. Based on this geometric topology, a CFD model was built to analyze the temperature distribution and obtain the effective thermal properties. The current numerical model was shown to be in good agreement with the existing experimental data for effective thermal conductivity available in the literature. supported by National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2015GB108002, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  12. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  13. Comparison of the pebbles of the Shinarump and Moss Back members of the Chinle formation

    USGS Publications Warehouse

    Albee, Howard Franklin

    1956-01-01

    Lithology, color, size, sphericity, and roundness of pebbles from the Shinarurnp and Moss Back members of the Chinle formation were analyzed and compared. The difference in the quartz:quartzite:chert ratios of the pebbles, the presence of limestone and siltstone pebbles, and to a lesser degree,the difference in color of pebbles serve to distinguish the Moss Back from the Shinarump. In areas where both the Moss Back and Shinarump are present, the average ratios of quartz, quartzite, a.nd chert are respectively about 12:37:51 and 82:16:2. Limestone and siltstone pebbles are commonly found in the Moss Back, whereas they are rarely found in the Shinarump. The colors of the Moss Back pebbles are generally darker than those of the Shinarump pebbles. The Moss Back contains more gray to black pebbles and fewer light-colored pebbles, such as red, orange, and white, than the Shinarump. Size, sphericity, and roundness of pebbles do not show a significant difference between the two units. Fossiliferous pebbles in the Moss Back and Shinarump were derived chiefly from sediments of Carboniferous and Permian ages and could have had common sources.

  14. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Daniels, F.

    1957-10-15

    Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

  15. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  16. Uranium assessment for the Precambrian pebble conglomerates in southeastern Wyoming

    SciTech Connect

    Borgman, L.E.; Sever, C.; Quimby, W.F.; Andrew, M.E.; Karlstrom, K.E.; Houston, R.S.

    1981-03-01

    This volume is a geostatistical resource estimate of uranium and thorium in quartz-pebble conglomerates, and is a companion to Volume 1: The Geology and Uranium Potential to Precambrian Conglomerates in the Medicine Bow Mountains and Sierra Madre of Southeastern Wyoming; and to Volume 2: Drill-Hole Data, Drill-Site Geology, and Geochemical Data from the Study of Precambrian Uraniferous Conglomerates of the Medicine Bow Mountains and the Sierra Madre of Southeastern Wyoming.

  17. Pebble orientation on large, experimental debris-flow deposits

    USGS Publications Warehouse

    Major, J.J.

    1998-01-01

    Replicable, pronounced orientation of discoid pebbles (??? 8 mm) embedded on surfaces of large (??? 10 m3) experimental debris-flow deposits reveals that strongly aligned, imbricate fabric can develop rapidly over short distances in mass flows. Pebble long axes aligned subparallel to deposit margins as well as subparallel to margins of surge waves arrested within the deposits. Pebble alignment exhibited modes both parallel to (a(p)), and normal to (a(t)), the primary flow direction; intermediate axes dipped preferentially inward from surge-wave margins (b(i) orientation). Repetitive development of margin-parallel, imbricate fabric distributed across deposit surfaces provides compelling evidence that deposits formed dominantly through progressive incremental accretion rather than through simple en masse emplacement. Pronounced fabric along deposit and arrested surge-wave margins reflects significant grain interaction along flow margins. This sedimentological evidence for significant marginal grain interaction complements theoretical analyses (Iverson, 1997) and other experimental data (Major, 1996: Iverson, 1997) that indicate that resistance along flow margins is an important factor affecting debris-flow deposition. The fabric on the experimental deposits demonstrates that debris flows can develop strongly imbricate particle orientation that mimics fabric developed during fluvial deposition. Particle shape and local stress fields appear to have more control over fabric development than does general depositional process. Other criteria in addition to particle orientation are needed to discriminate mass flow from fluvial gravel deposits and to unravel depositional history. ?? 1998 Elsevier Science B.V. All rights reserved.

  18. Uncertainty and Sensitivity Analyses of a Pebble Bed HTGR Loss of Cooling Event

    DOE PAGES

    Strydom, Gerhard

    2013-01-01

    The Very High Temperature Reactor Methods Development group at the Idaho National Laboratory identified the need for a defensible and systematic uncertainty and sensitivity approach in 2009. This paper summarizes the results of an uncertainty and sensitivity quantification investigation performed with the SUSA code, utilizing the International Atomic Energy Agency CRP 5 Pebble Bed Modular Reactor benchmark and the INL code suite PEBBED-THERMIX. Eight model input parameters were selected for inclusion in this study, and after the input parameters variations and probability density functions were specified, a total of 800 steady state and depressurized loss of forced cooling (DLOFC) transientmore » PEBBED-THERMIX calculations were performed. The six data sets were statistically analyzed to determine the 5% and 95% DLOFC peak fuel temperature tolerance intervals with 95% confidence levels. It was found that the uncertainties in the decay heat and graphite thermal conductivities were the most significant contributors to the propagated DLOFC peak fuel temperature uncertainty. No significant differences were observed between the results of Simple Random Sampling (SRS) or Latin Hypercube Sampling (LHS) data sets, and use of uniform or normal input parameter distributions also did not lead to any significant differences between these data sets.« less

  19. PUCs move to halt Pebble Springs, Limerick nukes

    SciTech Connect

    Not Available

    1982-06-01

    Public utility commission (PUC) opposition to nuclear-power-plant construction in Oregon and Pennsylvania indicates a new trend for PUCs to take the initiative against nuclear projects. By not allowing utilities to finance new plants with construction work in progress (CWIP) costs added to the rate base, the Pennsylvania PUC essentially cancelled the Limerick units in accordance with the sentiment of the state legislature. The Oregon PUC ordered Pacific Power and Light Co. to write off investments in two Pebble Springs units and retire the financial liability. Both issues will be settled in the courts. (DCK)

  20. HTR-Proteus Pebble Bed Experimental Program Cores 5,6,7,&8: Columnar Hexagonal Point-on-Point Packing with a 1:2 Moderator-to-Fuel Pebble Ratio

    SciTech Connect

    Bess, John D.; Sterbentz, James W.; Snoj, Luka; Lengar, Igor; Koberl, Oliver

    2015-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  1. HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

    SciTech Connect

    John D. Bess

    2013-03-01

    PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen critical configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.

  2. Stepped-anneal helium release in 1-mm beryllium pebbles from COBRA-1A2

    SciTech Connect

    Oliver, B.M.

    1998-03-01

    Stepped-anneal helium release measurements on two sets of fifteen beryllium pebbles irradiated in the Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-w), are reported. The purpose of the measurements was to determine the helium release characteristics of the beryllium using larger sample sizes and longer anneal times relative to earlier measurements. Sequential helium analyses were conducted over a narrower temperature range from approximately 800 C to 1100 C in 100 C increments, but with longer anneal time periods. To allow for overnight and unattended operation, a temperature controller and associated circuitry were added to the experimental setup. Observed helium release was nonlinear with time at each temperature interval, with each step being generally characterized by an initial release rate followed by a slowing of the rate over time. Sample Be-C03 showed a leveling off in the helium release after approximately 3 hours at a temperature of 890 C. Sample Be-D03, on the other hand, showed a leveling off only after {approximately}12 to 24 hours at a temperature of 1100 C. This trend is consistent with that observed in earlier measurements on single microspheres from the same two beryllium lots. None of the lower temperature steps showed any leveling off of the helium release. Relative to the total helium concentrations measured earlier, the total helium releases observed here represent approximately 80% and 92% of the estimated total helium in the C03 and D03 samples, respectively.

  3. Functional constraints on nest characteristics of pebble mounds of breeding male hornyhead chub Nocomis biguttatus.

    PubMed

    Wisenden, B D; Unruh, A; Morantes, A; Bury, S; Curry, B; Driscoll, R; Hussein, M; Markegard, S

    2009-11-01

    Breeding male hornyhead chub Nocomis biguttatus constructed nests in areas with relatively high but less than maximum flow rate and greater than average water depth. Nests comprised c. 3000 pebbles for a total mass of 11 kg. Males selected pebbles of smaller diameter but higher density than pebbles in the immediate vicinity. Thus, nests balanced the risk of mound erosion and energetic cost of nest construction with the benefits of protection from egg predators and a stable internal flow rate for oxygenation. These data help establish environmental management goals for the conservation of N. biguttatus and the lotic ecosystems dependent upon them.

  4. Forming Giant Planet Cores by Pebble Accretion -- Why Slow and Steady wins the Race

    NASA Astrophysics Data System (ADS)

    Kretke, Katherine A.; Levison, Harold F.

    2014-05-01

    In recent years there has been a radical new solution proposed to solve the problem of giant planet core formation. "Pebbles", particles ranging from centimeters to meters in size, have been shown to accrete extremely efficiently due to aerodynamic drag. Large capture cross-sections combined with fast pebble drift rates can allow a single planetesimal to grow from Ceres size to 10s of Earth masses well within the lifetime of gaseous circumstellar disks. However, at large sizes, the the capture-cross section of pebbles goes with the Hill sphere, forcing pebble accretion to becomes a fundamentally "oligarchic-like" process. This makes it difficult to form a few giant planet cores; instead a more generic result is many 10s to 100s of competing oligarchs. In this work, we present a way to get around this oligarchic dilemma If pebbles are assumed to form slowly over a long period of time, then the planetesimal growth rates are slow enough for the planetesimals to dynamically excite each other. As the larger planetisimals/proto-planets stir their smaller companions, these smaller bodies are excited to such a degree that they spend only a small fraction of their orbits embedded in the cooler pebble disk. This allows the larger bodies to starve their neighbors and maintain a relative runaway growth rate to high mass, effectively forming the cores of giant planets.

  5. Reactor Physics Characterization of the HTR Module with UCO Fuel

    SciTech Connect

    Gerhard Strydom

    2011-01-01

    ABSTRACT The HTR Module [1] is a graphite-moderated, helium cooled pebble bed High Temperature Reactor (HTR) design that has been extensively used as a reference template for the former South African and current Chinese HTR [2] programs. This design utilized spherical fuel elements packed into a dynamic pebble bed, consisting of TRISO coated uranium oxide (UO2) fuel kernels with a U-235 enrichment of 7.8% and a Heavy Metal loading of 7 grams per pebble. The main objective of this study is to compare several important reactor physics and core design parameters for the HTR Module and an identical design utilizing UCO fuel kernels. Fuel kernels of this type are currently being tested in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) as part of the larger Next Generation Nuclear Plant (NGNP) project. The PEBBED-THERMIX [3] code, which was developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two designs.

  6. The radial dependence of pebble accretion rates: A source of diversity in planetary systems. I. Analytical formulation

    NASA Astrophysics Data System (ADS)

    Ida, S.; Guillot, T.; Morbidelli, A.

    2016-06-01

    Context. The classical planetesimal accretion scenario for the formation of planets has recently evolved with the idea that pebbles, centimeter- to meter-sized icy grains migrating in protoplanetary disks, can control planetesimal and/or planetary growth. Aims: We investigate how pebble accretion depends on disk properties and affects the formation of planetary systems. Methods: We construct analytical models of pebble accretion onto planetary embryos that consistently account for the mass and orbital evolution of the pebble flow and reflect disk structure. Results: We derive simple formulas for pebble accretion rates in the so-called settling regime for planetary embryos that are more than 100 km in size. For relatively smaller embryos or in outer disk regions, the accretion mode is three-dimensional (3D), meaning that the thickness of the pebble flow must be taken into account, and resulting in an accretion rate that is independent of the embryo mass. For larger embryos or in inner regions, the accretion is in a two-dimensional (2D) mode, i.e., the pebble disk may be considered infinitely thin. We show that the radial dependence of the pebble accretion rate is different (even the sign of the power-law exponent changes) for different disk conditions such as the disk heating source (viscous heating or stellar irradiation), drag law (Stokes or Epstein, and weak or strong coupling), and in the 2D or 3D accretion modes. We also discuss the effect of the sublimation and destruction of icy pebbles inside the snow line. Conclusions: Pebble accretion easily produces a large diversity of planetary systems. In other words, to infer the results of planet formation through pebble accretion correctly, detailed prescriptions of disk evolution and pebble growth, sublimation, destruction and migration are required.

  7. Experimental Study and Computational Simulations of Key Pebble Bed Thermo-mechanics Issues for Design and Safety

    SciTech Connect

    Tokuhiro, Akira; Potirniche, Gabriel; Cogliati, Joshua; Ougouag, Abderrafi

    2014-07-08

    An experimental and computational study, consisting of modeling and simulation (M&S), of key thermal-mechanical issues affecting the design and safety of pebble-bed (PB) reactors was conducted. The objective was to broaden understanding and experimentally validate thermal-mechanic phenomena of nuclear grade graphite, specifically, spheres in frictional contact as anticipated in the bed under reactor relevant pressures and temperatures. The contact generates graphite dust particulates that can subsequently be transported into the flowing gaseous coolent. Under postulated depressurization transients and with the potential for leaked fission products to be adsorbed onto graphite 'dust', there is the potential for fission products to escape from the primary volume. This is a design safety concern. Furthermore, earlier safety assessment identified the distinct possibility for the dispersed dust to combust in contact with air if sufficient conditions are met. Both of these phenomena were noted as important to design review and containing uncertainty to warrant study. The team designed and conducted two separate effects tests to study and benchmark the potential dust-generation rate, as well as study the conditions under which a dust explosion may occure in a standardized, instrumented explosion chamber.

  8. Penn State geoPebble system: Design,Implementation, and Initial Results

    NASA Astrophysics Data System (ADS)

    Urbina, J. V.; Anandakrishnan, S.; Bilen, S. G.; Fleishman, A.; Burkett, P.

    2014-12-01

    The Penn State geoPebble system is a new network of wirelessly interconnected seismic and GPS sensor nodes with flexible architecture. This network will be used for studies of ice sheets in Antarctica and Greenland, as well as to investigate mountain glaciers. The network will consist of ˜150 geoPebbles that can be deployed in a user-defined spatial geometry. We present our design methodology, which has enabled us to develop these state-of- the art sensors using commercial-off-the-shelf hardware combined with custom-designed hardware and software. Each geoPebble is a self- contained, wirelessly connected sensor for collecting seismic measurements and position information. Key elements of each node encompasses a three-component seismic recorder, which includes an amplifier, filter, and 24- bit analog-to-digital converter that can sample up to 10 kHz. Each unit also includes a microphone channel to record the ground-coupled airwave. The timing for each node is available from GPS measurements and a local precision oscillator that is conditioned by the GPS timing pulses. In addition, we record the carrier-phase measurement of the L1 GPS signal in order to determine location at sub-decimeter accuracy (relative to other geoPebbles within a few kilometers radius). Each geoPebble includes 16 GB of solid-state storage, wireless communications capability to a central supervisory unit, and auxiliary measurements capability (including tilt from accelerometers, absolute orientation from magnetometers and temperature). A novel aspect of the geoPebble is a wireless charging system for the internal battery (using inductive coupling techniques). The geoPebbles include all the sensors (geophones, GPS, microphone), communications (WiFi), and power (battery and charging) internally, so the geoPebble system can operate without any cabling connections (though we do provide an external connector so that different geophones can be used). We report initial field-deployment results and

  9. Enhanced photoacoustic neuroimaging with gold nanorods and PEBBLEs

    NASA Astrophysics Data System (ADS)

    Witte, Russell S.; Kim, K.; Agarwal, A.; Fan, W.; Kopelman, R.; Kotov, N.; Kipke, D.; O'Donnell, M.

    2008-02-01

    Photoacoustic (PA) imaging provides excellent optical contrast with decent penetration and high spatial resolution, making it attractive for a variety of neural applications. We evaluated optical contrast agents with high absorption in the near infrared (NIR) as potential enhancers for PA neuroimaging: optical dyes, gold nanorods (GNRs) and PEBBLEs loaded with indocyanine green. Two PA systems were developed to test these agents in excised neural tissue and in vivo mouse brain. Lobster nerves were stained with the agents for 30 minutes and placed in a hybrid nerve chamber capable of electrical stimulation and recording, optical spectroscopy and PA imaging. Contrast agents boosted the PA signal by at least 30 dB using NIR illumination from a tunable pulsed laser. Photobleaching may be a limiting factor for optical dyes-the PA signal decreased steadily with laser illumination. The second setup enabled in vivo transcranial imaging of the mouse brain. A custom clinical ultrasound scanner and a 10-MHz linear array provided near real-time images during and after an injection of 2 nM gold nanorods into the tail vein. The peak PA signal from the brain vasculature was enhanced by up to 2 dB at 710 nm. Temporal dynamics of the PA signal were also consistent with mixing of the GNRs in the blood. These studies provide a baseline for enhanced PA imaging in neural tissue. The smart contrast agents employed in this study can be further engineered for molecular targeting and controlled drug delivery with potential treatment for a myriad of neural disorders.

  10. A panoptic model for planetesimal formation and pebble delivery

    NASA Astrophysics Data System (ADS)

    Krijt, S.; Ormel, C. W.; Dominik, C.; Tielens, A. G. G. M.

    2016-02-01

    Context. The journey from dust particle to planetesimal involves physical processes acting on scales ranging from micrometers (the sticking and restructuring of aggregates) to hundreds of astronomical units (the size of the turbulent protoplanetary nebula). Considering these processes simultaneously is essential when studying planetesimal formation. Aims: The goal of this work is to quantify where and when planetesimal formation can occur as the result of porous coagulation of icy grains and to understand how the process is influenced by the properties of the protoplanetary disk. Methods: We develop a novel, global, semi-analytical model for the evolution of the mass-dominating dust particles in a turbulent protoplanetary disk that takes into account the evolution of the dust surface density while preserving the essential characteristics of the porous coagulation process. This panoptic model is used to study the growth from sub-micron to planetesimal sizes in disks around Sun-like stars. Results: For highly porous ices, unaffected by collisional fragmentation and erosion, rapid growth to planetesimal sizes is possible in a zone stretching out to ~10 AU for massive disks. When porous coagulation is limited by erosive collisions, the formation of planetesimals through direct coagulation is not possible, but the creation of a large population of aggregates with Stokes numbers close to unity might trigger the streaming instability (SI). However, we find that reaching conditions necessary for SI is difficult and limited to dust-rich disks, (very) cold disks, or disks with weak turbulence. Conclusions: Behind the snow-line, porosity-driven aggregation of icy grains results in rapid (~104 yr) formation of planetesimals. If erosive collisions prevent this, SI might be triggered for specific disk conditions. The numerical approach introduced in this work is ideally suited for studying planetesimal formation and pebble delivery simultaneously and will help build a coherent

  11. Mechanics, kinematics and geometry of pebble abrasion from binary collisions

    NASA Astrophysics Data System (ADS)

    Miller, K. L.; Jerolmack, D. J.

    2014-12-01

    As sediment is transported downstream as bedload, it collides with the bed causing sharp edges to chip and wear away, rounding the rock through the process of abrasion. Previous work has linked abrasion to downstream fining and rounding of grains, however, there has been little attempt to understand the underlying kinematics of abrasion. Furthermore, most studies neglect the fine particle produced during the abrasion process, as the initial grain gets smaller and rounder. In this research, we preform well-controlled laboratory experiments to determine the functional dependence between impact energy and mass lost from abrasion. We use a double-pendulum "Newton's Cradle" set-up to examine the abrasion between two grains and with a high-speed camera, we can quantify the impact energies during collision. Results from experiments verify that mass loss is proportional to kinetic energy. We define a material parameter that incorporates material density, Young's modulus, and tensile stress and show that this parameter is directly related to the proportionality between mass loss and energy. We identify an initial region of the mass loss curves in which abrasion is independent of energy and material properties; results suggest this region is determined by shape. We show that grain size distributions of daughter products are universal and independent of material; they follow a Weibull distribution, which is expected distribution from brittle fracture theory. Finally, scanning electron microscope (SEM) images show a thin damage zone near the surface, suggesting that collision energy is attenuated over some small skin depth. Overall, we find that pebble abrasion by collision can be characterized by two universal scaling relations - the mass loss versus energy curves and the size distribution of daughter products. Results will be useful for estimating expected abrasion rates in the field, and additionally demonstrate that low-energy collisions produce large quantities of sand

  12. Comet 67P/Churyumov-Gerasimenko preserved the pebbles that formed planetesimals

    NASA Astrophysics Data System (ADS)

    Fulle, Marco; Della Corte, V.; Rotundi, A.; Rietmeijer, F. J. M.; Green, S. F.; Weissman, P.; Accolla, M.; Colangeli, L.; Ferrari, M.; Ivanovski, S.; Lopez-Moreno, J. J.; Epifani, E. Mazzotta; Morales, R.; Ortiz, J. L.; Palomba, E.; Palumbo, P.; Rodriguez, J.; Sordini, R.; Zakharov, V.

    2016-09-01

    Solar System formation models predict that the building-blocks of planetesimals were mm- to cm-sized pebbles, aggregates of ices and non-volatile materials, consistent with the compact particles ejected by comet 67P/Churyumov-Gerasimenko (67P hereafter) and detected by GIADA (Grain Impact Analyzer and Dust Accumulator) on-board the Rosetta spacecraft. Planetesimals were formed by the gentle gravitational accretion of pebbles, so that they have an internal macroporosity of 40%. We measure the average dust bulk density ρ _D = 795_{-65}^{+840} kg m-3 that, coupled to the 67P nucleus bulk density, provides the average dust-to-ices mass ratio δ = 8.5. We find that the measured densities of the 67P pebbles are consistent with a mixture of (15 ± 6)% of ices, (5 ± 2)% of Fe-sulfides, (28 ± 5)% of silicates, and (52 ± 12)% of hydrocarbons, in average volume abundances. This composition matches both the solar and CI-chondritic chemical abundances, thus showing that GIADA has sampled the typical non-volatile composition of the pebbles that formed all planetesimals. The GIADA data do not constrain the abundance of amorphous silicates vs. crystalline Mg,Fe-olivines and pyroxenes. We find that the pebbles have a microporosity of (52 ± 8)% (internal volume filling factor φP = 0.48 ± 0.08), implying an average porosity for the 67P nucleus of (71 ± 8)%, lower than previously estimated.

  13. The giant Pebble Cu-Au-Mo deposit and surrounding region, southwest Alaska: introduction

    USGS Publications Warehouse

    Kelley, Karen D.; Lang, James R.; Eppinger, Robert G.

    2013-01-01

    The Pebble deposit is located about 320 km southwest of and 27 km northwest of the village of Iliamna in Alaska (Fig. 1A). It is one of the largest porphyry deposits in terms of contained Cu (Fig. 2A) and it has the largest Au endowment of any porphyry deposit in the world (Fig. 2B). The deposit comprises the Pebble West and Pebble East zones that represent two coeval hydrothermal centers within a single system (Lang et al., 2013). Together the measured and indicated resources total 5,942 million metric tons (Mt) at 0.42% Cu, 0.35 g/t Au, and 250 ppm Mo with an inferred resource of 4,835 Mt at 0.24% Cu, 0.26 g/t Au, and 215 ppm Mo. In addition, the deposit contains significant concentrations of Ag, Pd, and Re (Northern Dynasty Minerals, 2011).

  14. Challenges in forming the solar system's giant planet cores via pebble accretion

    SciTech Connect

    Kretke, K. A.; Levison, H. F.

    2014-12-01

    Though ∼10 M {sub ⊕} mass rocky/icy cores are commonly held as a prerequisite for the formation of gas giants, theoretical models still struggle to explain how these embryos can form within the lifetimes of gaseous circumstellar disks. In recent years, aerodynamic-aided accretion of 'pebbles', objects ranging from centimeters to meters in size, has been suggested as a potential solution to this long-standing problem. While pebble accretion has been demonstrated to be extremely effective in local simulations that look at the detailed behavior of these pebbles in the vicinity of a single planetary embryo, to date there have been no global simulations demonstrating the effectiveness of pebble accretion in a more complicated, multi-planet environment. Therefore, we have incorporated the aerodynamic-aided accretion physics into LIPAD, a Lagrangian code that can follow the collisional/accretional/dynamical evolution of a protoplanetary system, to investigate how pebble accretion manifests itself in the larger planet formation picture. We find that under generic circumstances, pebble accretion naturally leads to an 'oligarchic' type of growth in which a large number of planetesimals grow to similar-sized planets. In particular, our simulations tend to form hundreds of Mars- and Earth-mass objects between 4 and 10 AU. While merging of some oligarchs may grow massive enough to form giant planet cores, leftover oligarchs lead to planetary systems that cannot be consistent with our own solar system. We investigate various ideas presented in the literature (including evaporation fronts and planet traps) and find that none easily overcome this tendency toward oligarchic growth.

  15. Pebble shale (Early Cretaceous) depositional environments in National Petroleum Reserve in Alaska (NPRA)

    SciTech Connect

    Blanchard, D.C.; Tailleur, I.L.

    1983-03-01

    A pebble shale of Lower Cretaceous age occurs across the North Slope and continues into northwestern Canada. This organic shale (1 to 5% organic carbon) is possibly the source for the Prudhoe Bay hydrocarbons and includes localized well-developed sand bodies such as those in the giant Kuparuk oil field. The inferred rifting of the Arctic basin, subsequent subsidence of a northern source area, and the southern orogeny during the Late Jurassic and Early Cretaceous contributed to the unique lithology and regional setting of the pebble shale. Overlying the uppermost unconformity on the Barrow arch a pebbly mudstone 3 to 8-m (10 to 26 ft) thick, of Hauterivian-Barremian age contains well-rounded sand grains, pebbles and cobbles, pelletal glaunconite, shell fragments, wood chips, and burrows. The pebbly mudstone facies, derived from a northerly source, is interpreted as a lowstand and subaqueous delta environment formed as the northern provenance was uplifted during Valanginian time. Rapid subsidence of this basin margin was related to the inferred Atlantic margin type downfaulting north of the Barrow arch. The subsequent deposits produced a blanket of sediment 300,000 mi/sup 2/ (482,700 km/sup 2/) in areal extent. This (2 to 9 ft, 0.8 m), fine-grained sands occur within the pebble shale of the Barremian silty shale facies and are restricted to the Barrow area. These sands have an average combined thickness of 45 ft (15 m) and contain traces of oil. Paleogeographic considerations imply better reservoir and coalescing of these sands to the north toward the paleoshoreline and suggest potential for discovery of hydrocarbon reserves from drilling locations on the northern barrier island systems and offshore.

  16. The Transient 3-D Transport Coupled Code TORT-TD/ATTICA3D for High-Fidelity Pebble-Bed HTGR Analyses

    NASA Astrophysics Data System (ADS)

    Seubert, Armin; Sureda, Antonio; Lapins, Janis; Bader, Johannes; Laurien, Eckart

    2012-01-01

    This article describes the 3D discrete ordinates-based coupled code system TORT-TD/ATTICA3D that aims at steady state and transient analyses of pebble-bed high-temperature gas cooled reactors. In view of increasing computing power, the application of time-dependent neutron transport methods becomes feasible for best estimate evaluations of safety margins. The calculation capabilities of TORT-TD/ATTICA3D are presented along with the coupling approach, with focus on the time-dependent neutron transport features of TORT-TD. Results obtained for the OECD/NEA/NSC PBMR-400 benchmark demonstrate the transient capabilities of TORT-TD/ATTICA3D.

  17. Growing Pebbles and Conceptual Prisms - Understanding the Source of Student Misconceptions about Rock Formation.

    ERIC Educational Resources Information Center

    Kusnick, Judi

    2002-01-01

    Analyzes narrative essays--stories of rock formation--written by pre-service elementary school teachers. Reports startling misconceptions among preservice teachers on pebbles that grow, human involvement in rock formation, and sedimentary rocks forming as puddles as dry up, even though these students had completed a college level course on Earth…

  18. Catalyst functionalized buffer sorbent pebbles for rapid separation of carbon dioxide from gas mixtures

    SciTech Connect

    Aines, Roger D

    2015-03-31

    A method for separating CO.sub.2 from gas mixtures uses a slurried media impregnated with buffer compounds and coating the solid media with a catalyst or enzyme that promotes the transformation of CO.sub.2 to carbonic acid. Buffer sorbent pebbles with a catalyst or enzyme coating are provided for rapid separation of CO.sub.2 from gas mixtures.

  19. Thermal ramp tritium release in COBRA-1A2 C03 beryllium pebbles

    SciTech Connect

    Baldwin, D.L.

    1998-03-01

    Tritium release kinetics, using the method of thermal ramp heating at three linear ramp rates, were measured on the COBRA-1A2 C03 1-mm beryllium pebbles. This report includes a brief discussion of the test, and the test data in graph format.

  20. The impact of ellipsoidal particle shape on pebble breakage in gravel

    PubMed Central

    Tuitz, Christoph; Exner, Ulrike; Frehner, Marcel; Grasemann, Bernhard

    2012-01-01

    We have studied the influence of particle shape and consequently loading configuration on the breakage load of fluvial pebbles. Unfortunately, physical strength tests on pebbles, i.e., point-load tests, can only be conducted under one specific stable loading configuration. Therefore, the physical uniaxial strength tests performed in this study were extended by a two-dimensional finite-element stress analysis, which is capable of investigating those scenarios that are not possible in physical tests. Breakage load, equivalent to that measured in unidirectional physical tests, was determined from the results of the stress analysis by a maximum tensile stress-based failure criterion. Using this assumption, allows the determination of breakage load for a range of different kind of synthetic loading configurations and its comparison with the natural breakage load distribution of the physical strength tests. The results of numerical modelling indicated that the configuration that required the least breakage load corresponded with the minor principal axis of the ellipsoidal pebbles. In addition, most of the simulated gravel-hosted loading configurations exceeded the natural breakage load distribution of fluvial pebbles obtained from the physical strength tests. PMID:26321870

  1. Fabrication development and preliminary characterization of Li 2TiO 3 pebbles by wet process

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Fuchinoue, Katsuhiro; Sawada, Hiroshi; Watarumi, Kazutoshi

    1998-10-01

    Lithium titanate (Li 2TiO 3) has attracted attention of many researchers because of easy tritium recovery at low temperature, high chemical stability, etc. The application of small Li 2TiO 3 spheres has been proposed in some designs of fusion blanket. Although, the wet process and sol-gel method are the most advantageous as a fabrication method of Li 2TiO 3 pebbles from points of mass production, and of reprocessing necessary for effective use of resources and reduction of radioactive wastes. However, the fabrication of Li 2TiO 3 pebbles by the wet process has not been established. Therefore, in this study, fabrication development and preliminary characterization of Li 2TiO 3 pebbles by the wet process were performed, noting the aging and sintering conditions in the fabrication process of gel-spheres. At the best condition, Li 2TiO 3 pebbles with the target density of 80-85%T.D. were obtained.

  2. Catalyst functionalized buffer sorbent pebbles for rapid separation of carbon dioxide from gas mixtures

    DOEpatents

    Aines, Roger D.

    2013-03-12

    A method for separating CO.sub.2 from gas mixtures uses a slurried media impregnated with buffer compounds and coating the solid media with a catalyst or enzyme that promotes the transformation of CO.sub.2 to carbonic acid. Buffer sorbent pebbles with a catalyst or enzyme coating are provided for rapid separation of CO.sub.2 from gas mixtures.

  3. On the water delivery to terrestrial embryos by ice pebble accretion

    NASA Astrophysics Data System (ADS)

    Sato, Takao; Okuzumi, Satoshi; Ida, Shigeru

    2016-05-01

    Standard accretion disk models suggest that the snow line in the solar nebula migrated interior to the Earth's orbit in a late stage of nebula evolution. In this late stage, a significant amount of ice could have been delivered to 1 AU from outer regions in the form of mm to dm-sized pebbles. This raises the question why the present Earth is so depleted of water (with the ocean mass being as small as 0.023% of the Earth mass). Here we quantify the amount of icy pebbles accreted by terrestrial embryos after the migration of the snow line assuming that no mechanism halts the pebble flow in outer disk regions. We use a simplified version of the coagulation equation to calculate the formation and radial inward drift of icy pebbles in a protoplanetary disk. The pebble accretion cross section of an embryo is calculated using analytic expressions presented by recent studies. We find that the final mass and water content of terrestrial embryos strongly depends on the radial extent of the gas disk, the strength of disk turbulence, and the time at which the snow lines arrives at 1 AU. The disk's radial extent sets the lifetime of the pebble flow, while turbulence determines the density of pebbles at the midplane where the embryos reside. We find that the final water content of the embryos falls below 0.023 wt% only if the disk is compact (<100 AU), turbulence is strong at 1 AU, and the snow line arrives at 1 AU later than 2-4 Myr after disk formation. If the solar nebula extended to 300 AU, initially rocky embryos would have evolved into icy planets of 1-10 Earth masses unless the snow-line migration was slow. If the proto-Earth contained water of ~1 wt% as might be suggested by the density deficit of the Earth's outer core, the formation of the proto-Earth was possible with weaker turbulence and with earlier (>0.5-2 Myr) snow-line migration.

  4. Implications of pebble accretion on the composition of hot and cold Jupiters

    NASA Astrophysics Data System (ADS)

    Bitsch, Bertram; Johansen, Anders; Madhusudhan, Nikku

    2016-10-01

    The formation of the planetary cores of gas giants via the accretion of planetesimals takes very long and is not compatible with the lifetime of protoplanetary discs (Levison et al. 2010). This time-scale problem can be solved through the accretion of pebbles onto a planetary seed. Contrary to planetesimals, pebbles feel the headwind from the gas which robs them of angular momentum allowing an efficient growth from the entire Hill sphere, which reduces the growth time-scale by several orders of magnitude (Lambrechts & Johansen, 2012; 2014). However, pebble accretion self-terminates when the planets start to open a partial gap in the disc, which accelerates the gas outside of the planets orbit to super-Keplerian speeds and thus stops the flow of pebbles onto the planetary core (Lambrechts et al. 2014). Typically this mass is of the order of 10-20 Earth masses, depending on the local disc properties. The planet can then start to accrete a gaseous envelope without a pollution of pebbles. During its growth, the planet migrates through the disc, which evolves in time (Bitsch et al. 2015a,b).Different volatile species like CO2 or H2O have different condensation temperatures and are thus present in either solid or gaseous form at different locations in the disc. A pebble accreting planet can thus only accrete volatiles that are in solid form, while a gas accreting planet will only accrete volatiles which are in gaseous form. Therefore the final chemical composition of the planetary atmosphere of a giant planet is strongly influenced by the formation location of the initial planetary seed and its subsequent migration path through the disc. Additionally, the envelope can be enriched through the erosion of the planetary core.I will discuss the implications of the formation of planets via pebble accretion and their subsequent migration through the disc on the composition of gas giants. In particular I will focus on the carbon to oxygen ratio of hot Jupiters around other stars

  5. Thermo-mechanical and neutron lifetime modelling and design of Be pebbles in the neutron multiplier for the LIFE engine

    NASA Astrophysics Data System (ADS)

    DeMange, P.; Marian, J.; Caro, M.; Caro, A.

    2009-11-01

    Concept designs for the laser inertial fusion/fission energy (LIFE) engine include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a reliable and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermo-mechanical behaviour under continued neutron exposure. We consider the effects of high fluence and fast fluxes on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 °C to enable creep to relax the stresses induced by swelling. Under these circumstances, we estimate the pebble lifetime to be at least 16 months if uncoated, and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  6. Growing the gas-giant planets by the gradual accumulation of pebbles.

    PubMed

    Levison, Harold F; Kretke, Katherine A; Duncan, Martin J

    2015-08-20

    It is widely held that the first step in forming gas-giant planets, such as Jupiter and Saturn, was the production of solid 'cores' each with a mass roughly ten times that of the Earth. Getting the cores to form before the solar nebula dissipates (in about one to ten million years; ref. 3) has been a major challenge for planet formation models. Recently models have emerged in which 'pebbles' (centimetre-to-metre-sized objects) are first concentrated by aerodynamic drag and then gravitationally collapse to form objects 100 to 1,000 kilometres in size. These 'planetesimals' can then efficiently accrete left-over pebbles and directly form the cores of giant planets. This model is known as 'pebble accretion'; theoretically, it can produce cores of ten Earth masses in only a few thousand years. Unfortunately, full simulations of this process show that, rather than creating a few such cores, it produces a population of hundreds of Earth-mass objects that are inconsistent with the structure of the Solar System. Here we report that this difficulty can be overcome if pebbles form slowly enough to allow the planetesimals to gravitationally interact with one another. In this situation, the largest planetesimals have time to scatter their smaller siblings out of the disk of pebbles, thereby stifling their growth. Our models show that, for a large and physically reasonable region of parameter space, this typically leads to the formation of one to four gas giants between 5 and 15 astronomical units from the Sun, in agreement with the observed structure of the Solar System.

  7. Growing the gas-giant planets by the gradual accumulation of pebbles.

    PubMed

    Levison, Harold F; Kretke, Katherine A; Duncan, Martin J

    2015-08-20

    It is widely held that the first step in forming gas-giant planets, such as Jupiter and Saturn, was the production of solid 'cores' each with a mass roughly ten times that of the Earth. Getting the cores to form before the solar nebula dissipates (in about one to ten million years; ref. 3) has been a major challenge for planet formation models. Recently models have emerged in which 'pebbles' (centimetre-to-metre-sized objects) are first concentrated by aerodynamic drag and then gravitationally collapse to form objects 100 to 1,000 kilometres in size. These 'planetesimals' can then efficiently accrete left-over pebbles and directly form the cores of giant planets. This model is known as 'pebble accretion'; theoretically, it can produce cores of ten Earth masses in only a few thousand years. Unfortunately, full simulations of this process show that, rather than creating a few such cores, it produces a population of hundreds of Earth-mass objects that are inconsistent with the structure of the Solar System. Here we report that this difficulty can be overcome if pebbles form slowly enough to allow the planetesimals to gravitationally interact with one another. In this situation, the largest planetesimals have time to scatter their smaller siblings out of the disk of pebbles, thereby stifling their growth. Our models show that, for a large and physically reasonable region of parameter space, this typically leads to the formation of one to four gas giants between 5 and 15 astronomical units from the Sun, in agreement with the observed structure of the Solar System. PMID:26289203

  8. Uncovering East Antarctic Bedrock using detrital zircon geochronology and pebble lithologies from Mount Howe, Scott Glacier

    NASA Astrophysics Data System (ADS)

    Dits, T.; Licht, K.; Bader, N.; Kaplan, M. R.; Schaefer, J. M.; Winckler, G.

    2012-12-01

    Till from the flanks of Mount Howe, the southernmost outcrop in the world at the head of the Scott Glacier, Antarctica, offers an exclusive view of East Antarctic bedrock through analysis of detrital zircon geochronology and pebble lithology. With no outcrops upstream of the Mount Howe nunatak, detrital zircons and pebbles incorporated in the supraglacial till place direct new age and lithologic constraints on unmapped, ice covered bedrock in the Scott Glacier catchment. Nine moraine crests were sampled along a 2 km transect from the modern ice edge toward exposed Beacon Supergroup bedrock, where rock weathering increases away from the ice margin. Preliminary cosmogenic ages on boulders on the same crests as the provenance study indicate most of the moraine complex formed over the last 100 ka, but some ridges close to the headwall may be much older. Pebble lithologies across the transect show minimal statistical variation, averaging 60% mafic igneous, 30% metamorphic, and 10% sedimentary lithologies dominantly from the Ferrar and Beacon Supergroups. Observations of faceting and striations on pebble surfaces reveal that up to 40-50% of the pebble fraction of the till was subglacially transported, and a minimum of 15% are exotic lithologies. Nearly 80% of cobbles collected from a non-random survey reveal the presence of several exotic rock types, including vesicular olivine basalt, quartzite, and four different compositions of granite. Guided by backscatter electron imagery of detrital zircons, 385 ages from U-Pb isotopes of detrital zircons from 8 sequential moraine crests were determined by laser ablation-inductively coupled plasma mass spectroscopy (LA-ICPMS). Distinct age populations were identified at 185-190 Ma, 255-270 Ma, 355-365 Ma, 550-580 Ma, and 2740 Ma. Four samples in the middle of the transect all display a similar 1010-1040 Ma peak that is statistically different from the remaining samples. The 185 Ma population differs from the typical East Antarctic

  9. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    SciTech Connect

    Trond Bjornard; John Hockert

    2011-08-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work summarized in

  10. Fabrication of Li2TiO3 pebbles using PVA-boric acid reaction for solid breeding materials

    NASA Astrophysics Data System (ADS)

    Park, Yi-Hyun; Cho, Seungyon; Ahn, Mu-Young

    2014-12-01

    Lithium metatitanate (Li2TiO3) is a candidate breeding material of the Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM). The breeding material is used in pebble-bed form to reduce the uncertainty of the interface thermal conductance. In this study, Li2TiO3 pebbles were successfully fabricated by the slurry droplet wetting method using the cross-linking reaction between polyvinyl alcohol (PVA) and boric acid. The effects of fabrication parameters on the shaping of Li2TiO3 green body were investigated. In addition, the basic characteristics of the sintered pebble were also evaluated. The shape of Li2TiO3 green bodies was affected by slurry viscosity, PVA content and boric acid content. The grain size and average crush load of sintered Li2TiO3 pebble were controlled by the sintering time. The boron was completely removed during the final sintering process.

  11. Studies on crude oil removal from pebbles by the application of biodiesel.

    PubMed

    Xia, Wen-xiang; Xia, Yan; Li, Jin-cheng; Zhang, Dan-feng; Zhou, Qing; Wang, Xin-ping

    2015-02-15

    Oil residues along shorelines are hard to remove after an oil spill. The effect of biodiesel to eliminate crude oil from pebbles alone and in combination with petroleum degrading bacteria was investigated in simulated systems. Adding biodiesel made oil detach from pebbles and formed oil-biodiesel mixtures, most of which remained on top of seawater. The total petroleum hydrocarbon (TPH) removal efficiency increased with biodiesel quantities but the magnitude of augment decreased gradually. When used with petroleum degrading bacteria, the addition of biodiesel (BD), nutrients (NUT) and BD+NUT increased the dehydrogenase activity and decreased the biodegradation half lives. When BD and NUT were replenished at the same time, the TPH removal efficiency was 7.4% higher compared to the total improvement of efficiency when BD and NUT was added separately, indicating an additive effect of biodiesel and nutrients on oil biodegradation.

  12. Sojourner Rover View of Well-Rounded Pebbles in Cabbage Patch

    NASA Technical Reports Server (NTRS)

    1997-01-01

    Sojourner Rover image of rounded 4-cm-wide pebble (lower center) and excavation of cloddy deposit of Cabbage Patch at lower left. Note the bright wind tails of drift material extending from small rocks and the wheel track from upper right to lower left.

    Well-rounded objects, like the one in this image, were not seen at the Viking sites. These are thought to be pebbles liberated from sedimentary rocks composed of cemented silts, sands and rounded fragments; such rocks are called conglomerates.

    NOTE: original caption as published in Science Magazine

    Mars Pathfinder is the second in NASA's Discovery program of low-cost spacecraft with highly focused science goals. The Jet Propulsion Laboratory, Pasadena, CA, developed and manages the Mars Pathfinder mission for NASA's Office of Space Science, Washington, D.C. JPL is a division of the California Institute of Technology (Caltech).

  13. The dynamical evolution of the asteroid belt in the pebble accretion scenario

    NASA Astrophysics Data System (ADS)

    Pirani, Simona; Mustill, Alexander; Turrini, Diego; Johansen, Anders

    2016-10-01

    The high excitation of the asteroid belt could be the trace of a past cohexistence of asteroids and planetary embryos. After the formation of Jupiter and Saturn, the asteroid belt lost about 99% of its mass, depleted by gravitational interactions with these giant planets and it was left with only Ceres as a relic of the planetary embryo population. Our aim is to construct a main belt (based on new estimates for the birth distribution of asteroids and planetary embryos that grow by pebble accretion) and test its evolution with different parameters and configurations of the giant planets. We test new pebble accretion growth tracks for the giant planets and compare the evolution of the asteroid belt to the classical in-situ growth.

  14. BRILLIANT PEBBLES: A METHOD FOR DETECTION OF VERY LARGE INTERSTELLAR GRAINS

    SciTech Connect

    Socrates, Aristotle; Draine, Bruce T. E-mail: draine@astro.princeton.edu

    2009-09-01

    A photon of wavelength {lambda} {approx} 1 {mu}m interacting with a dust grain of radius a{sub p} {approx} 1 mm (a 'pebble') undergoes scattering in the forward direction, largely within a small characteristic diffraction angle {theta}{sub s} {approx} {lambda}/a{sub p} {approx} 100''. Though millimeter-size dust grains contribute negligibly to the interstellar medium's visual extinction, the signal they produce in scattered light may be detectable, especially for variable sources. Observations of light scattered at small angles allow for the direct measurement of the large grain population; variable sources can also yield tomographic information of the interstellar medium's mass distribution. The ability to detect brilliant pebble halos requires a careful understanding of the instrument point-spread function.

  15. Thermo-mechanical and neutron lifetime modeling and design of Be pebbles in the neutron multiplier for the LIFE engine

    SciTech Connect

    DeMange, P; Marian, J; de Caro, M S; Caro, A

    2009-03-16

    Concept designs for the laser-initiated fusion/fission engine (LIFE) include a neutron multiplication blanket containing Be pebbles flowing in a molten salt coolant. These pebbles must be designed to withstand the extreme irradiation and temperature conditions in the blanket to enable a safe and cost-effective operation of LIFE. In this work, we develop design criteria for spherical Be pebbles on the basis of their thermomechanical behavior under continued neutron exposure. We consider the effects of high fluence/fast flux on the elastic, thermal and mechanical properties of nuclear-grade Be. Our results suggest a maximum pebble diameter of 30 mm to avoid tensile failure, coated with an anti-corrosive, high-strength metallic shell to avoid failure by pebble contact. Moreover, we find that the operation temperature must always be kept above 450 C to enable creep to relax the stresses induced by swelling, which we estimate to be at least 16 months if uncoated and up to six years when coated. We identify the sources of uncertainty on the properties used and discuss the advantages of new intermetallic beryllides and their use in LIFE's neutron multiplier. To establish Be-pebble lifetimes with improved confidence, reliable experiments to measure irradiation creep must be performed.

  16. Fabrication and characterization of Li2TiO3 core-shell pebbles with enhanced lithium density

    NASA Astrophysics Data System (ADS)

    Hong, Ming; Zhang, Yingchun; Mi, Yingying; Xiang, Maoqiao; Zhang, Yun

    2014-02-01

    In order to increase the lithium density and control the lithium mass loss at elevated temperature, development of Li2TiO3 pebbles with excess Li is needed. In this paper, Li2TiO3 core-shell pebbles with different Li/Ti molar ratios were fabricated by a gel-casting method using Li2TiO3 and Li2CO3 as starting materials. Differential thermal analysis appending a thermogravimetric analyzer (DTA-TG) and X-ray diffraction (XRD) were employed to understand the solid-state reactions. And then the calcining and sintering processes were optimized. Microstructure, element distribution, crush load and density of the pebbles were also investigated. The experimental results showed that the pebble had a Li2TiO3-Li4TiO4 complex phase core and a tunable thickness Li2TiO3 shell, and the lithium density of the pebbles significantly increased with the increasing of the Li/Ti ratio. The optimum Li/Ti ratio was 2.7, and the pebbles displayed a good crush load (about 32 N) when sintered at 950 °C for 2 h in N2 atmosphere.

  17. Alexa Fluor 488 as an iron sensing molecule and its application in PEBBLE nanosensors.

    PubMed

    Sumner, James P; Kopelman, Raoul

    2005-04-01

    Molecular Probes' Alexa Fluor dyes are generally used for biological labeling because of their ideal fluorescent properties, but here we detail Alexa Fluor 488's nanomolar sensitivity to free iron. Furthermore, the dye has been encapsulated into a polymer nanosphere by a microemulsion method, producing <100 nm particles. These nanosensors, PEBBLEs (Probe Encapsulated By Biologically Localized Embedding) have micromolar sensitivity and are non-responsive to other metal ions of biological interest.

  18. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    SciTech Connect

    Ilas, Germina; Ilas, Dan; Kelly, Ryan P; Sunny, Eva E

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  19. Validation of In-Situ Iron-Manganese Oxide Coated Stream Pebbles as Sensors for Arsenic Source Monitoring

    NASA Astrophysics Data System (ADS)

    Blake, J.; Peters, S. C.; Casteel, A.

    2013-12-01

    Locating nonpoint source contaminant fluxes can be challenging due to the inherent heterogeneity of source and of the subsurface. Contaminants such as arsenic are a concern for drinking water quality and ecosystem health. Arsenic contamination can be the result of several natural and anthropogenic sources, and therefore it can be difficult to trace and identify major areas of arsenic in natural systems. Identifying a useful source indicator for arsenic is a crucial step for environmental remediation efforts. Previous studies have found iron-manganese oxide coated streambed pebbles as useful source indicators due to their high attraction for heavy metals in water. In this study, pebbles, surface water at baseflow and nearby rocks were sampled from the Pennypack Creek and its tributaries, in southwestern Pennsylvania, to test the ability of coated streambed pebbles as environmental source indicators for arsenic. Quartz pebbles, 5-7 cm in diameter, were sampled to minimize elemental contamination from rock chemistry. In addition, quartz provides an excellent substrate for iron and manganese coatings to form. These coatings were leached from pebbles using 4M nitric acid with 0.1% concentrated hydrochloric acid. Following sample processing, analyses were performed using an ICP-MS and the resulting data were spatially organized using ArcGIS software. Arsenic, iron and manganese concentrations in the leachate are normalized to pebble surface area and each location is reported as a ratio of arsenic to iron and manganese. Results suggest that iron-manganese coated stream pebbles are useful indicators of arsenic location within a watershed.

  20. Accretion of Cometary Nuclei in the Solar Nebula: Boulders, Not Pebbles

    NASA Astrophysics Data System (ADS)

    Weissman, Paul R.; A'Hearn, Michael

    2015-11-01

    Comets are the most primitive bodies in the solar system. They retain a largely unprocessed record of conditions in the primordial solar nebula 4.56 Gyr ago, including the initial accretion of dust and ice particles into macroscopic bodies. Current accretion theory suggests that ice and dust aggregates grew to pebble (cm) sizes before streaming instabilities and gravitational collapse brought these pebble swarms together as km-sized (or larger) bodies. Recent imaging of the nucleus of comet 67P/Churyumov-Gerasimenko by the Rosetta OSIRIS camera team has revealed the existence of “goose bump” terrain on the nucleus surface and lining the interior walls of large, ~200 m diameter and 180 m deep cylindrical pits. These pits are believed to be sinkholes, formed when near-surface materials collapse into voids within the nucleus, revealing the fresh comet interior on the walls of the pits. The goose bump terrain consists of 3-4 m diameter “boulders” randomly stacked one on top of another. We propose that these boulders, likely with an icy-conglomerate composition, are the basic building blocks of cometary nuclei. This is the first observational confirmation of current accretion theories, with the caveat that rather than pebbles, the preferred size range is 3-4 m boulders for objects formed in the giant planets region of the solar system. The presence of icy grains beyond the solar nebula snow-line and the large heliocentric range of the giant planets region likely contribute to the formation of these larger boulders, before they are incorporated into cometary nuclei. This work was supported by NASA through the U.S. Rosetta Project.

  1. Making Planet Nine: Pebble Accretion at 250-750 AU in a Gravitationally Unstable Ring

    NASA Astrophysics Data System (ADS)

    Kenyon, Scott J.; Bromley, Benjamin C.

    2016-07-01

    We investigate the formation of icy super-Earth mass planets within a gravitationally unstable ring of solids orbiting at 250-750 AU around a 1 {M}⊙ star. Coagulation calculations demonstrate that a system of a few large oligarchs and a swarm of pebbles generates a super-Earth within 100-200 Myr at 250 AU and within 1-2 Gyr at 750 AU. Systems with more than ten oligarchs fail to yield super-Earths over the age of the solar system. As these systems evolve, destructive collisions produce detectable debris disks with luminosities of {10}-5{--}{10}-3 relative to the central star.

  2. Upper and Lower I/O Bounds for Pebbling r-Pyramids

    NASA Astrophysics Data System (ADS)

    Ranjan, Desh; Savage, John; Zubair, Mohammad

    Modern computers have several levels of memory hierarchy. To obtain good performance on these processors it is necessary to design algorithms that minimize I/O traffic to slower memories in the hierarchy. In this paper, we present I/O efficient algorithms to pebble r-pyramids and derive lower bounds on the number of I/O steps to do so. The r-pyramid graph models financial applications which are of practical interest and where minimizing memory traffic can have a significant impact on cost saving.

  3. Making Planet Nine: Pebble Accretion at 250–750 AU in a Gravitationally Unstable Ring

    NASA Astrophysics Data System (ADS)

    Kenyon, Scott J.; Bromley, Benjamin C.

    2016-07-01

    We investigate the formation of icy super-Earth mass planets within a gravitationally unstable ring of solids orbiting at 250–750 AU around a 1 {M}ȯ star. Coagulation calculations demonstrate that a system of a few large oligarchs and a swarm of pebbles generates a super-Earth within 100–200 Myr at 250 AU and within 1–2 Gyr at 750 AU. Systems with more than ten oligarchs fail to yield super-Earths over the age of the solar system. As these systems evolve, destructive collisions produce detectable debris disks with luminosities of {10}-5{--}{10}-3 relative to the central star.

  4. Provenance of the Subinal Formation, Central Guatemala, Based on Point-Counting of Pebbles in Conglomerates

    NASA Astrophysics Data System (ADS)

    Gutierrez, A.; Martens, U.

    2007-05-01

    The continental Subinal Formation of Central Guatemala is composed of red conglomerates and sandstones that outcrop in the Motagua valley between the San Agustín and Cabañas fault, and in the southeastern corner of Guatemala. Stratigraphic position implies a post-Cretaceous depositional age, which has not been accurately established. Point counts of pebbles in conglomerates were performed in localities distributed along the Motagua valley from Granados to Los Amates, and in the Jocotán-Esquipulas area near the border with Honduras. Pebble types contained in conglomerates in the Motagua area are very diverse, and include sandstones, limestones, chert, milky quartz, phyllite, marble, chlorite schist, quartz-muscovite schist, amphibolite, eclogite, gneiss, granite, gabbro/diorite, volcanic rocks, and abundant serpentinite. Schist and gneiss resembles rocks of the Chuacús complex. Serpentinite, gabbro, amphibolite, and eclogite clasts are similar to oceanic lithosphere from the Motagua suture. Abundant white mica in interbedded sandstones is ubiquitous, suggesting derivation from the mica-rich Chuacús complex. Unfoliated granites and volcanic clasts were probably derived from the northern edge of the Chortis block or an unknown terrane, as no unmetamorphosed igneous rocks are known from the southern edge of the Maya block. The relative abundance of some clast groups correlates with the rock units exposed immediately adjacent to the north across the San Agustín fault. Serpentinite in conglomerate is most abundant near Juan de Paz; eclogite, amphibolite and gneiss are most abundant in the Granados area; and white mica is most abundant where Chuacús complex schists are most pelitic. This suggests that the Cabañas fault accommodated most of the strike-slip movement of the Motagua fault system. In contrast, conglomerates in southeastern Guatemala are more homogeneous, containing chiefly clasts of volcanic origin, with minor limestone, graphitic phyllite, chert, and

  5. Code System for Reactor Physics and Fuel Cycle Simulation.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterativemore » processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.« less

  6. Code System for Reactor Physics and Fuel Cycle Simulation.

    SciTech Connect

    TEUCHERT, E.

    1999-04-21

    Version 00 VSOP94 (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shut-down features, in-core and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. In addition to its use in research and development work for the High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors.

  7. Geology and uranium evaluation of the precambrian quartz-pebble conglomerates of the Needle Mountains, Southwest Colorado. Final report

    SciTech Connect

    Burns, L.K.; Ethridge, F.G.; Tyler, N.; Gross, A.S.; Campo, A.M.

    1980-05-01

    Precambrian quartz-pebble conglomerates of the Needle Mountains of southwest Colorado are present in the Vallecito Conglomerate and the basal conglomerate of Uncompahgre Formation. Three conglomerates have lithologic, sedimentologic and mineralogic characteristics that are similar to those of the Precambrian uranium-bearing fossil placer deposits of the Witwatersrand in South Africa and the Blind River Elliot Lake district in Ontario, Canada. However, the Precambrian quartz-pebble conglomerates of Colorado do not contain the key indicator minerals of primary detrital pyrite, uraninite, and other uranium mineral species. The absence of these minerals and the structural and stratigraphic relations of these two formations suggest that they were deposited between 1780 and 1460 million years ago. These formations are, therefore, not old enough to have been deposited under the reducing atmospheric conditions thought to be necessary for the formation of placer uranium deposits. The average uranium content of the Precambrian quartz-pebble conglomerates sampled is 2.9 ppM based on analyses of 502 rock samples. Maximum values for uranium (total by neutron activation are 11 ppM for the Vallecito Conglomerate and 5.0 ppM for the basal part of the Uncompahgre Formation. The Vallecito Conglomerate contains on the average 1.2 ppM total uranium and the Uncompahgre Formation contains 1.2 ppM total uranium. These low values further indicate the low potential of these quartz-pebble conglomerates as fossil-placer uranium deposit. Uranium contents of up to 330 ppM were found in dark slates of the Uncompahgre Formation; however, these units were not closely associated with the pebble conglomerates. Some potential uranium targets of vein type and in black pelitic units are found in the Uncompahgre Formation, but these are not the fossil-placer quartz-pebble conglomerate type of deposits.

  8. Transport of marked pebbles in short periods of time on a coarse clastic beach (Marina di Pisa, Italy)

    NASA Astrophysics Data System (ADS)

    Bertoni, D.; Ciavola, P.; Grottoli, E.; Sarti, G.

    2012-04-01

    Transport of coarse sediments on coarse clastic beaches still presents aspects that are not fully understood. For instance, there is a generally perceived notion that during fair-weather periods coarse grains hardly move, if not at all. The aim of this experiment is to prove that sediments such as pebbles are subject to significant shift in very short lapses of time and under low energy waves. An artificial coarse clastic beach at Marina di Pisa (Tuscany, Italy) was chosen as study site: Barbarossa beach is 110 m long and is bounded by two groynes. The mean grain size is about 40-to-50 mm. About 80 pebbles were marked by means of the RFID technology, which enables to univocally identify the tracers. The marked pebbles were released along cross-shore transects (one pebble each on the fair-weather berm, on the beachface and on the step crest) on the morning of September 15th, and two recovery campaigns were carried out after 6 and 24 hours from the injection. No particular wave activity was recorded during the time frame of the experiment. After the first recovery campaign, which was performed 6 hours later than the injection, about 94% of the pebbles were detected. After the second recovery campaign, 24 hours later, the recovery rate decreased to 89%. Considering that the technique provides for detection of tracers within 50 cm, the resulting loss of pebbles after so brief spans of time is remarkable. The lack of detection of few tracers implies that the transport rate that they experienced is not negligible. The highest rate of losses was recorded on the beachface, the zone that is subjected the most to waves even under calm conditions. Pebble movement is also confirmed by the fact that tracers detected after the first recovery campaign were not detected once again after the second recovery campaign, and vice versa. The results of the experiment are useful to better define the transport of coarse sediments, verifying that pebbles have to be expected be moving even

  9. A preliminary study on removal of AMD precipitate coatings on pebbles

    NASA Astrophysics Data System (ADS)

    Lee, W.; Min, K.; Lee, H.

    2011-12-01

    AMD(acid mine drainage) having a low pH and elevated concentrations of heavy metals affects environments as a major pollutant. In addition to AMD's water contamination, reddish brown precipitates from AMD spoil the watercourse scenery without suitable removal treatments. To examine the removal potentiality of ultrasonic cleaner, the pebble samples coated by reddish brown precipitates were collected at abandoned mine stream and scraped precipitate coatings were analyzed for their chemical compositions and mineralogy. Their average contents of Fe2O3, SO3, and Al2O3 were 84.3%, 6.13%, and 3.69%, respectively and goethite was the major constituent mineral. Laboratorial tests to remove precipitate coatings were performed in an ultrasonic cleaner with the frequency of 40kHz at 20 to 70oC for 10 to 60 minutes. Water and hydrochloric acid of 0.1M to 1M were used as a cleaning solvent and the ratio of solvent to precipitate coated pebbles was 5 in weight. In result, an ultrasonic cleaning treatment is expected to be applied successively in field and removal efficiency was increased as reaction time, temperature, and concentration of solvent rises.

  10. Building the giant planet cores by convergent migration of pebble-accreting embryos

    NASA Astrophysics Data System (ADS)

    Chrenko, Ondrej; Broz, Miroslav

    2016-10-01

    An explanation of the accretion buildup of giant planet cores on rather short (~Myr) time scales remains a long-standing challenge for scenarios of planetary system formation. One of the recently proposed processes that can take part during this evolutionary stage is the convergent Type I migration of Earth-sized embryos towards the zero-torque radius, occurring at an opacity transition within the dusty-gaseous protoplanetary disk (e.g. Pierens et al. 2013). Inconveniently, simulations show that such groups of embryos do not merge easily because they often get locked in mutual mean-motion resonances and consequently form an inward-migrating convoy.We revise this possibility of merging embryos while taking into account their ongoing growth by pebble accretion. Our aim is to check whether the rapid changes of masses combined with the migration of embryos through the feeding zone can break the resonant chain and allow for the giant planet core formation.The environment of the protoplanetary disk is modeled with the 2D FARGO code (Masset 2000), which we modified in order to perform non-isothermal hydrodynamic simulations, assuming flux-limited radiative diffusion (Levermore & Pomraning 1981). The embedded massive bodies are evolved simultaneously in 3D using the hybrid Wisdom-Holman/Gauss-Radau integrator from the Rebound package (Rein & Spiegel 2015). A semi-analytic method is used to evolve the masses of embryos by pebble accretion (e.g. Levison et al. 2015).

  11. Evidence of fast pebble growth near condensation fronts in the HL Tau protoplanetary disk

    NASA Astrophysics Data System (ADS)

    Zhang, Ke; Blake, Geoffrey; Bergin, Edwin

    2015-08-01

    Water and simple organic molecular ice dominates the mass of solid materials available for planetesimal and planet formation beyond the water snow line. Around the condensation fronts of these most abundant volatile species, rapid pebble growth is predicted to occur based on recent numerical simulations of dust coagulation and settling. Here we propose that the dips in interferometric images of the HL Tau protoplanetary disk are due to rapid pebble growth around condensation fronts. HL Tau, a young star in Taurus molecular cloud, was observed by ALMA at 110, 233 and 343GHz continuum bands with spatial resolution as good as 3 AU. We show that the three dips at distances of 13, 32 and 63 AU are spatially resolved and that their center radii are coincident with the expected mid-plane condensation fronts of water ice, ammonia or water-ammonia hydrates, and water-clatherates (with CO2, methane, CO and N2). The 63 AU dip is much wider than the first two and the full extent of the dip also covers the condensation front of pure CO2. The spectral index map of HL Tau between 233 and 343 GHz shows that the flux ratios inside the dip regions are statistically larger than that of nearby regions in the disk. The variation can be explained by a model with two dust populations where most of dust mass resides in the population which has grown into decimeter size scales inside the dips.

  12. Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

    NASA Astrophysics Data System (ADS)

    Kawamura, H.; Ishitsuka, E.; Tsuchiya, K.; Nakamichi, M.; Uchida, M.; Yamada, H.; Nakamura, K.; Ito, H.; Nakazawa, T.; Takahashi, H.; Tanaka, S.; Yoshida, N.; Kato, S.; Ito, Y.

    2003-08-01

    The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

  13. Detrital zircon U-Pb geochronology and provenance of the Carboniferous-Permian glaciomarine pebbly slates in the Tibetan Plateau

    NASA Astrophysics Data System (ADS)

    Wang, Q.; Zhu, D.; Zhao, Z.; Chung, S.; Li, C.; Sui, Q.; Fu, X.; Mo, X.

    2011-12-01

    Glaciomarine diamictites (including pebbly slate, pebbly siltstone, and pebbly sandstone) in the Tibetan Plateau are widely interpreted to have been associated with the deglaciation of the Indian continent. Guiding by zircon cathodoluminescence images, we determined U-Pb ages for detrital zircons from five typical Carboniferous-Permian pebbly slate samples from the Qiangtang, Lhasa, and Tethyan Himalaya of the Tibetan Plateau. The age distributions of detrital zircons from two samples (180 analyses) from Qiwu and Gangma Tso of the Qiangtang Terrane are similar, with two main age peaks ca. 579 and ca. 816 Ma and one minor age peak ca. 2490 Ma. Two samples (177 analyses) from Jiangrang and Damxung of the Lhasa Terrane define similar age distributions with two main age peaks ca. 539 and ca. 1175 Ma. Ages of detrital zircons from one sample (110 analyses) from Kangmar of the Tethyan Himalaya display main age peaks ca. 535, ca. 949, and ca. 2490 Ma. The ca. 816-Ma detrital zircons from the Qiangtang Terrane were most likely derived from the Lesser Himalaya, and the ca. 950-Ma detrital zircons from the Tethyan Himalaya might have been sourced from the High Himalaya, Eastern Ghats Province of the Indian plate and the Rayner Province of East Antarctica. The distinctive ca. 1175-Ma age population characteristic of zircons in the pebbly slates from the Lhasa Terrane is identical to the detrital zircons from the late Paleozoic sandstones (Zhu et al., 2011a) and the inherited zircons from the Mesozoic peraluminous granites (Zhu et al., 2011b) in this terrane, but significantly absent in the pebbly slates from both the Qiangtang and the Tethyan Himalayan terranes. The ca. 1175-Ma detrital zircons in the Lhasa Terrane were most likely sourced from the Albany-Fraser-Wilkes in southwestern Australia and East Antarctica. These new data obtained in this study reveal a distinct difference of detrital zircon provenance for the coeval Carboniferous-Permian glaciomarine pebbly slates

  14. Using pebble lithology and roundness to interpret gravel provenance in piedmont fluvial systems of the Rocky Mountains, USA

    USGS Publications Warehouse

    Lindsey, D.A.; Langer, W.H.; Van Gosen, B. S.

    2007-01-01

    Clast populations in piedmont fluvial systems are products of complex histories that complicate provenance interpretation. Although pebble counts of lithology are widely used, the information provided by a pebble count has been filtered by a potentially large number of processes and circumstances. Counts of pebble lithology and roundness together offer more power than lithology alone for the interpretation of provenance. In this study we analyze pebble counts of lithology and roundness in two contrasting fluvial systems of Pleistocene age to see how provenance varies with drainage size. The two systems are 1) a group of small high-gradient incised streams that formed alluvial fans and terraces and 2) a piedmont river that formed terraces in response to climate-driven cycles of aggradation and incision. We first analyze the data from these systems within their geographic and geologic context. After this is done, we employ contingency table analysis to complete the interpretation of pebble provenance. Small tributary streams that drain rugged mountains on both sides of the Santa Cruz River, southeast Arizona, deposited gravel in fan and terrace deposits of Pleistocene age. Volcanic, plutonic and, to a lesser extent, sedimentary rocks are the predominant pebble lithologies. Large contrasts in gravel lithology are evident among adjacent fans. Subangular to subrounded pebbles predominate. Contingency table analysis shows that hard volcanic rocks tend to remain angular and, even though transport distances have been short, soft tuff and sedimentary rocks tend to become rounded. The Wind River, a major piedmont stream in Wyoming, drains rugged mountains surrounding the northwest part of the Wind River basin. Under the influence of climate change and glaciation during the Pleistocene, the river deposited an extensive series of terrace gravels. In contrast to Santa Cruz tributary gravel, most of the Wind River gravel is relatively homogenous in lithology and is rounded to

  15. Fabrication and improvement of the density of Li 2TiO 3 pebbles by the optimization of a sol-gel method

    NASA Astrophysics Data System (ADS)

    Wu, Xiangwei; Wen, Zhaoyin; Xu, Xiaogang; Han, Jinduo; Lin, Bin

    2009-08-01

    Li 2TiO 3 is regarded as a promising candidate breeder in solid blanket concepts. Pebble configuration has been a preferred option due to its potential advantages in blanket design. Li 2TiO 3 pebbles were successfully fabricated by a water-based sol-gel method previously. However, the sintered density of the pebbles was very low (less than 70% theoretical density). The process parameters were optimized and the sintered density of the pebbles was improved significantly in this work. Li 2TiO 3 pebbles with density as high as 85% were obtained at a relatively lower sintering temperature (1100 °C) and shorter sintering time (4 h). The experimental results showed that the viscosity of the sol was influential to the sphericity of the gel-spheres and thus the sintered pebbles. The variety of lithium source, the pH value of the solution and the sintering conditions demonstrated significant influences on the microstructure and density of the sintered Li 2TiO 3 pebbles.

  16. Formation and accumulation of radiation-induced defects and radiolysis products in modified lithium orthosilicate pebbles with additions of titanium dioxide

    NASA Astrophysics Data System (ADS)

    Zarins, Arturs; Valtenbergs, Oskars; Kizane, Gunta; Supe, Arnis; Knitter, Regina; Kolb, Matthias H. H.; Leys, Oliver; Baumane, Larisa; Conka, Davis

    2016-03-01

    Lithium orthosilicate (Li4SiO4) pebbles with 2.5 wt.% excess of silicon dioxide (SiO2) are the European Union's designated reference tritium breeding ceramics for the Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM). However, the latest irradiation experiments showed that the reference Li4SiO4 pebbles may crack and form fragments under operation conditions as expected in the HCPB TBM. Therefore, it has been suggested to change the chemical composition of the reference Li4SiO4 pebbles and to add titanium dioxide (TiO2), to obtain lithium metatitanate (Li2TiO3) as a second phase. The aim of this research was to investigate the formation and accumulation of radiation-induced defects (RD) and radiolysis products (RP) in the modified Li4SiO4 pebbles with different contents of TiO2 for the first time, in order to estimate and compare radiation stability. The reference and the modified Li4SiO4 pebbles were irradiated with accelerated electrons (E = 5 MeV) up to 5000 MGy absorbed dose at 300-990 K in a dry argon atmosphere. By using electron spin resonance (ESR) spectroscopy it was determined that in the modified Li4SiO4 pebbles, several paramagnetic RD and RP are formed and accumulated, like, E' centres (SiO33-/TiO33-), HC2 centres (SiO43-/TiO3-) etc. On the basis of the obtained results, it is concluded that the modified Li4SiO4 pebbles with TiO2 additions have comparable radiation stability with the reference pebbles.

  17. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  18. An orientation soil survey at the Pebble Cu-Au-Mo porphyry deposit, Alaska

    USGS Publications Warehouse

    Smith, Steven M.; Eppinger, Robert G.; Fey, David L.; Kelley, Karen D.; Giles, S.A.

    2009-01-01

    Soil samples were collected in 2007 and 2008 along three traverses across the giant Pebble Cu-Au-Mo porphyry deposit. Within each soil pit, four subsamples were collected following recommended protocols for each of ten commonly-used and proprietary leach/digestion techniques. The significance of geochemical patterns generated by these techniques was classified by visual inspection of plots showing individual element concentration by each analytical method along the 2007 traverse. A simple matrix by element versus method, populated with a value based on the significance classification, provides a method for ranking the utility of methods and elements at this deposit. The interpretation of a complex multi-element dataset derived from multiple analytical techniques is challenging. An example of vanadium results from a single leach technique is used to illustrate the several possible interpretations of the data.

  19. Rings, igloos, and pebbles of salt formed by drying saline drops.

    PubMed

    Shin, Bongsu; Moon, Myoung-Woon; Kim, Ho-Young

    2014-11-01

    It is well-known that evaporation of sessile drops with suspended particles like colloids and coffee powders can yield a variety of two-dimensional patterns depending on the particle shapes and internal flow patterns. Here we show that ordered three-dimensional structures can be built via evaporation of saline drops on highly hydrophobic substrates like pristine PP (polypropylene) with micropores and nanostructured low-surface-energy PP. On pristine PP having a high contact angle but a large contact angle hysteresis (CAH) with water, either rings or igloos of salt are formed depending on the salt concentration and evaporation rate. On nanostructured low-surface-energy PP having extreme water repellency with a very low CAH, pebbles of salt are formed regardless of salt concentration and evaporation rate. These observations lead us to conclude that combined effects of solubility, evaporation rate, and mobility of the contact line determine the final three-dimensional shape of the salt precipitate. PMID:25289755

  20. Uncertainty Analysis for a De-pressurised Loss of Forced Cooling Event of the PBMR Reactor

    SciTech Connect

    Jansen van Rensburg, Pieter A.; Sage, Martin G.

    2006-07-01

    This paper presents an uncertainty analysis for a De-pressurised Loss of Forced Cooling (DLOFC) event that was performed with the systems CFD (Computational Fluid Dynamics) code Flownex for the PBMR reactor. An uncertainty analysis was performed to determine the variation in maximum fuel, core barrel and reactor pressure vessel (RPV) temperature due to variations in model input parameters. Some of the input parameters that were varied are: thermo-physical properties of helium and the various solid materials, decay heat, neutron and gamma heating, pebble bed pressure loss, pebble bed Nusselt number and pebble bed bypass flows. The Flownex model of the PBMR reactor is a 2-dimensional axisymmetrical model. It is simplified in terms of geometry and some other input values. However, it is believed that the model adequately indicates the effect of changes in certain input parameters on the fuel temperature and other components during a DLOFC event. Firstly, a sensitivity study was performed where input variables were varied individually according to predefined uncertainty ranges and the results were sorted according to the effect on maximum fuel temperature. In the sensitivity study, only seven variables had a significant effect on the maximum fuel temperature (greater that 5 deg. C). The most significant are power distribution profile, decay heat, reflector properties and effective pebble bed conductivity. Secondly, Monte Carlo analyses were performed in which twenty variables were varied simultaneously within predefined uncertainty ranges. For a one-tailed 95% confidence level, the conservatism that should be added to the best estimate calculation of the maximum fuel temperature for a DLOFC was determined as 53 deg. C. This value will probably increase after some model refinements in the future. Flownex was found to be a valuable tool for uncertainly analyses, facilitating both sensitivity studies and Monte Carlo analyses. (authors)

  1. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect

    L.E. Demick

    2010-08-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  2. Basis for NGNP Reactor Design Down-Selection

    SciTech Connect

    L.E. Demick

    2011-11-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  3. Gas dynamics and heat transfer in a packed pebble-bed reactor for the 4th generation nuclear energy

    NASA Astrophysics Data System (ADS)

    Abdulmohsin, Rahman

    For over three decades, the presence of magnetic flux noise with a power spectral density scaling roughly as S phi ( f) ∝ 1/falpha where a≲1, has been known to limit the low-frequency performance of dc superconducting quantum interference devices (SQUIDs). In recent years, experiments indicate that this same noise persists to frequencies up to 1 GHz and is a dominant source of dephasing in flux-sensitive superconducting quantum bits (qubits). Thus, the reduction of flux noise presents a major hurdle towards the successful realization of scalable quantum computers that are based on flux-based qubits. In this thesis, we present experimental measurements, theoretical analyses, and numerical simulations that support a more detailed understanding of both the microscopic and macroscopic properties of flux n. Our experimental work begins with flux noise measurements of a large number of SQUIDs in the temperature range from 0.1 K to 4 K. We report on measurements of ten SQUIDs with systematically varied geometries and show that alpha increases as the temperature is lowered; in so doing, each spectrum pivots about a nearly constant frequency. The mean square flux noise, inferred by integrating the power spectra, grows rapidly with temperature and at a given temperature is approximately independent of the outer dimension of a given SQUID washer. We show that these results are incompatible with a model based on the random reversal of independent, spins that are located at the surface of the SQUID washer. In the course of our flux noise measurements, we became aware of a spurious contribution to low-frequency critical current noise in Josephson junctions normally attributed to charge trapping in the barrier arising from temperature instabilities inherent in cryogenic systems. These temperature fluctuations modify the critical current via its temperature dependence. By computing cross-correlations between measured temperature and critical current noise in Al-AlOx-Al junctions, we show that, despite excellent temperature stability, temperature fluctuations induce observable critical current fluctuations. Particularly, becuase 1/ f critical current noise has decreased with improved fabrication techniques in recent years, it is important to understand and eliminate this additional noise source. Next, we introduce a numerical method of calculating the mean square flux noise F2 from independently fluctuating spins on the surface of thin-film loops of arbitrary geometry. By reciprocity, F2 is proportional to Br2 , where B(r) is the magnetic field generated by a circulating current around the loop and r varies over the loop surface. By discretizing the loop nonuniformly, we efficiently and accurately compute the current distribution and resulting magnetic field, which may vary rapidly across the loop. We use this method to compute F2 in a number of scenarios in which we systematically vary physical parameters of the loop. We compare our simulations to an earlier analytic result predicting that F2 ∝ R/W in the limit where the loop radius R is much greater than the linewidth W. We further show that the previously neglected contribution of edge spins to F2 is significant---even dominant---in narrow-linewidth loops. To calculate theoretical dephasing rates in qubits, we consider flux noise with a spectral density Sphi( f) = A2/ (f/1 Hz) alpha, where A is of the order of 1 muphi 0 Hz--1/2 and 0.6 ≤ alpha ≤ 1.2; applied flux, our calculations of the dependence of the pure dephasing time tau φ Ramsey and echo pulse sequences on alpha for fixed A show that tauφ decreases rapidly as alpha is reduced. We find that tauφ is relatively insensitive to the noise bandwidth, f1 ≤ f ≤ f2 for all alpha provided the ultraviolet cutoff frequency f2 > 1/tauφ. We calculate the ratio tauφ,E/tau φ, R of the echo (E) and Ramsey (R) sequences, and the dependence of the decay function on alpha and f2. We investigate the case in which S phi(f0) is fixed at the "pivot frequency" f0 ≠ 1 Hz while alpha is varied, and find that the choice of f 0 can greatly influence the sensitivity of tauφ, E and tauφ, R to the value of alpha. Finally, we conclude with a brief review of our principal results and conclusions. We also comment on promising avenues of future research.

  4. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  5. Neutronics analyses in support of the conceptual design of the MAPS NTP reactor

    SciTech Connect

    Raepsaet, X.; Lenain, R.

    1996-03-01

    Within the framework of the French nuclear thermal propulsion program called MAPS (Lenain 1996), several neutronics studies and analyses were performed. The aim was to determine the basic design features of a reactor based on the Pebble Bed Reactor concept (Powell 1985) and needing minimum technological developments. In the concern to further enhance the reactor safety posture and to maintain a minimum engine mass breakdown, a beryllium moderated/reflected reactor using highly enriched UO{sub 2} or UC{sub 2} as fuel has been designed with a mean hydrogen core outlet temperature of 2200 K (theoretical ISP of 859 s). The objective of this paper is to give a detailed neutronics analysis of the MAPS reactor. {copyright} {ital 1996 American Institute of Physics.}

  6. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    SciTech Connect

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  7. USGS exploration geochemistry studies at the Pebble porphyry Cu-Au-Mo deposit, Alaska-pdf of presentation

    USGS Publications Warehouse

    Eppinger, Robert G.; Kelley, Karen D.; Fey, David L.; Giles, Stuart A.; Minsley, Burke J.; Smith, Steven M.

    2010-01-01

    From 2007 through 2010, scientists in the U.S. Geological Survey (USGS) have been conducting exploration-oriented geochemical and geophysical studies in the region surrounding the giant Pebble porphyry Cu-Au-Mo deposit in southwestern Alaska. The Cretaceous Pebble deposit is concealed under tundra, glacial till, and Tertiary cover rocks, and is undisturbed except for numerous exploration drill holes. These USGS studies are part of a nation-wide research project on evaluating and detecting concealed mineral resources. This report focuses on exploration geochemistry and comprises illustrations and associated notes that were presented as a case study in a workshop on this topic. The workshop, organized by L.G. Closs and R. Glanzman, is called 'Geochemistry in Mineral Exploration and Development,' presented by the Society of Economic Geologists at a technical conference entitled 'The Challenge of Finding New Mineral Resources: Global Metallogeny, Integrative Exploration and New Discoveries,' held at Keystone, Colorado, October 2-5, 2010.

  8. Geochemical Data for Samples Collected in 2007 Near the Concealed Pebble Porphyry Cu-Au-Mo Deposit, Southwest Alaska

    USGS Publications Warehouse

    Fey, David L.; Granitto, Matthew; Giles, Stuart A.; Smith, Steven M.; Eppinger, Robert G.; Kelley, Karen D.

    2008-01-01

    In the summer of 2007, the U.S. Geological Survey (USGS) began an exploration geochemical research study over the Pebble porphyry copper-gold-molydenum (Cu-Au-Mo) deposit in southwest Alaska. The Pebble deposit is extremely large and is almost entirely concealed by tundra, glacial deposits, and post-Cretaceous volcanic and volcaniclastic rocks. The deposit is presently being explored by Northern Dynasty Minerals, Ltd., and Anglo-American LLC. The USGS undertakes unbiased, broad-scale mineral resource assessments of government lands to provide Congress and citizens with information on national mineral endowment. Research on known deposits is also done to refine and better constrain methods and deposit models for the mineral resource assessments. The Pebble deposit was chosen for this study because it is concealed by surficial cover rocks, it is relatively undisturbed (except for exploration company drill holes), it is a large mineral system, and it is fairly well constrained at depth by the drill hole geology and geochemistry. The goals of the USGS study are (1) to determine whether the concealed deposit can be detected with surface samples, (2) to better understand the processes of metal migration from the deposit to the surface, and (3) to test and develop methods for assessing mineral resources in similar concealed terrains. This report presents analytical results for geochemical samples collected in 2007 from the Pebble deposit and surrounding environs. The analytical data are presented digitally both as an integrated Microsoft 2003 Access? database and as Microsoft 2003 Excel? files. The Pebble deposit is located in southwestern Alaska on state lands about 30 km (18 mi) northwest of the village of Illiamna and 320 km (200 mi) southwest of Anchorage (fig. 1). Elevations in the Pebble area range from 287 m (940 ft) at Frying Pan Lake just south of the deposit to 1146 m (3760 ft) on Kaskanak Mountain about 5 km (5 mi) to the west. The deposit is in an area of

  9. Granulometry of pebble beach ridges in Fort Williams Point, Greenwich Island, Antarctic Peninsula; a possible result from Holocene climate fluctuations

    USGS Publications Warehouse

    Santana, E.; Dumont, J.F.

    2007-01-01

    We present a granulometric study of emerged pebble beach ridges in the Fort Williams Point, Greenwich Island, Antarctic Peninsula. We studied 8 beach ridges from the shore up to 13.5 m above current sea level. The beach ridges are made of volcanic material from the surrounding relief, but also include glacially transported gneiss and granodiorite pebble and cobble. Based on granulometric distribution analysis of 2100 samples from 39 locations we identified evidence of 4 sequences of 1 to 3 ridges. Most of the material seems to be reworked from a till. Pavement formation by iceberg between the sequences of beach ridges suggests periods of lower temperature. The interpretation suggests that sequences of beach ridge construction formed during warmer periods of the late Holocene. This occurs in the framework of an isostatic postglacial uplift allowing the progressive mobilization of periglaciar material.

  10. Additives affecting properties of β-Li2TiO3 pebbles in a modified indirect wet chemistry process

    NASA Astrophysics Data System (ADS)

    Yu, Cheng-Long; Liu, Wei; Yang, Long-Tao; Wang, Dao-Yi; Wu, Kang; Zhang, Zeng-Ping; Wang, Xiu-Feng; Yanagisawa, Kazumichi

    2016-11-01

    Lithium metatitanate (β-Li2TiO3) pebbles were fabricated via the modified indirect wet chemistry method. Effect of varied additives, as polyvinyl alcohol, glycerol, and agar on the properties evolution was investigated. The highest density is obtained by adding 2 wt% (weight percent) polyvinyl alcohol, 3 wt% glycerol, and 3 wt% agar, respectively. β-Li2TiO3 pebbles with relative sintered density of 92.4%T.D. (Theoretical Density), the ratio of the intensity of diffraction peak (002) to that of (-133) of about 2.93, about 1.58 mm in diameter, a better sphericity of 1.02, the particle size of 5-6 μm, and the well-developed surface layered structure are successfully fabricated with 3 wt% glycerol. Glycerol is beneficial to improving the properties by other fabrication method as well.

  11. Pebbly mudstones in the Cretaceous Pigeon Point Formation, western California: a study in the transitional stages from submarine slumps to cohesive debris flows

    NASA Astrophysics Data System (ADS)

    López-Gamundí, Oscar R.

    1993-04-01

    The pebbly mudstones in the Late Cretaceous Pigeon Point Formation originated by slumping and related debris-flow processes in a submarine canyon/slope depositional system. The sedimentary characteristics of the pebbly mudstones (PM) enable the distinction of two main varieties: (a) heterogeneous or "patchy" pebbly mudstones (PPM) exhibiting irregular bed geometries and diffuse to irregular bed contacts, with maximum clast sizes in intraformational boulder-sized population, including abundant rip-up mudstone and sandstone clasts with common soft sediment deformations; (b) homogeneous pebbly mudstones (HPM) with tabular bed geometries, non-erosive and almost flat bed contacts, maximum clast sizes in extraformational pebble-sized fraction and scarce to absent soft-sediment deformations. The two varieties of pebbly mudstone represent the mechanical transition from slumps to cohesive debris flows. The presence of abundant intraformational clasts and disrupted, yet preserved slump-fold features in the PPM suggest that this facies represents a stage closer to the slump end-member. As the shear-strain progressed and a fully remolded cohesive debris flow developed, an almost complete disaggregation of the poorly consolidated sand and mud clasts and the incorporation into the remolded "matrix" phase took place.

  12. Fuel Development For Gas-Cooled Fast Reactors

    SciTech Connect

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  13. Fuel development for gas-cooled fast reactors

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  14. Tritium release kinetics from Li 2TiO 3 pebbles as prepared by soft-wet-chemistry

    NASA Astrophysics Data System (ADS)

    Casadio, S.; van der Laan, J. G.; Alvani, C.; Magielsen, A. J.; Stijkel, M. P.

    2004-08-01

    Lithium meta titanate pebbles has been prepared from agglomeration-sintering powders which were obtained by Li-Ti-peroxo-complex solution precursor (Li 2TiO 3 dissolved at room temperature in H 2O + 40% H 2O 2 and stabilized with citric acid). Through this wet route Li 2TiO 3 pebbles with high density(˜92% of T.D.) has been obtained and the tritium release behavior has been tested `in-pile' by the EXOTIC-8.9 experiment (˜440 days of irradiation at full power in the high neutron flux of HFR-Petten). Tritium residence times ( τ) in the pebbles has been measured during irradiation between 550 and 400 °C and He + 0.1%H 2 purge gas composition. By a thermally activated process (activation energy=111 kJ/mol) with 410 °C as minimum temperature the tritium residence time is found to be about 1 day, which places this specimen in a good ranking position among those tested by the EXOTIC-series. A clear increase of the tritium release rate has been observed by increased H 2 concentration (up to 1%) in the He purge. Out-of-pile ramp-annealing tritium desorption (TPD) tests on short-time irradiated pebbles has been also performed by various devices and conditions. The kinetic parameters from the TPD investigation gave consistent results with those characterizing the equilibrium times of tritium release rate after the gas composition and temperature transients imposed on the specimen during the in-pile experiment.

  15. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

    A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

  16. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  17. Geochemical Data for Samples Collected in 2008 Near the Concealed Pebble Porphyry Cu-Au-Mo Deposit, Southwest Alaska

    USGS Publications Warehouse

    Fey, David L.; Granitto, Matthew; Giles, Stuart A.; Smith, Steven M.; Eppinger, Robert G.; Kelley, Karen D.

    2009-01-01

    In the summer of 2007, the U.S. Geological Survey (USGS) began an exploration geochemical research study over the Pebble porphyry copper-gold-molybdenum deposit. This report presents the analytical data collected in 2008. The Pebble deposit is world class in size, and is almost entirely concealed by tundra, glacial deposits, and post-Cretaceous volcanic rocks. The Pebble deposit was chosen for this study because it is concealed by surficial cover rocks, is relatively undisturbed (except for exploration company drill holes), is a large mineral system, and is fairly well-constrained at depth by the drill hole geology and geochemistry. The goals of this study are to 1) determine whether the concealed deposit can be detected with surface samples, 2) better understand the processes of metal migration from the deposit to the surface, and 3) test and develop methods for assessing mineral resources in similar concealed terrains. The analytical data are presented as an integrated Microsoft Access 2003 database and as separate Excel files.

  18. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  19. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  20. Applying Pebble-Rotating Game to Enhance the Robustness of DHTs

    PubMed Central

    Ren, LiYong; Nie, XiaoWen; Dong, YuChi

    2013-01-01

    Distributed hash tables (DHTs) are usually used in the open networking environment, where they are vulnerable to Sybil attacks. Pebble-Rotating Game (PRG) mixes the nodes of the honest and the adversarial randomly, and can resist the Sybil attack efficiently. However, the adversary may have some tricks to corrupt the rule of PRG. This paper proposes a set of mechanisms to make the rule of PRG be obliged to obey. A new joining node must ask the Certificate Authority (CA) for its signature and certificate, which records the complete process on how a node joins the network and obtains the legitimacy of the node. Then, to prevent the adversary from accumulating identifiers, any node can make use of the latest certificate to judge whether one identifier is expired with the help of the replacement property of RPG. This paper analyzes in details the number of expired certificates which are needed to store in every node, and gives asymptotic solution of this problem. The analysis and simulations show that the mean number of the certificates stored in each node are , where n is the size of the network. PMID:23776485

  1. Close-in planetesimal formation by pile-up of drifting pebbles

    NASA Astrophysics Data System (ADS)

    Drążkowska, J.; Alibert, Y.; Moore, B.

    2016-10-01

    Context. The consistency of planet formation models suffers from the disconnection between the regime of small and large bodies. This is primarily caused by so-called growth barriers: the direct growth of larger bodies is halted at centimetre-sized objects and particular conditions are required for the formation of larger, gravitationally bound planetesimals. Aims: We aim to connect models of dust evolution and planetesimal formation to identify regions of protoplanetary discs that are favourable for the formation of kilometre-sized bodies and the first planetary embryos. Methods: We combine semi-analytical models of viscous protoplanetary disc evolution, dust growth and drift including backreaction of the dust particles on the gas, and planetesimal formation via the streaming instability into one numerical code. We investigate how planetesimal formation is affected by the mass of the protoplanetary disc, its initial dust content, and the stickiness of dust aggregates. Results: We find that the dust growth and drift leads to a global redistribution of solids. The pile-up of pebbles in the inner disc provides local conditions where the streaming instability is effective. Planetesimals form in an annulus with its inner edge lying between 0.3 AU and 1 AU and its width ranging from 0.3 AU to 3 AU. The resulting surface density of planetesimals follows a radial profile that is much steeper than the initial disc profile. These results support formation of terrestrial planets in the solar system from a narrow annulus of planetesimals, which reproduces their peculiar mass ratios.

  2. EVIDENCE OF FAST PEBBLE GROWTH NEAR CONDENSATION FRONTS IN THE HL TAU PROTOPLANETARY DISK

    SciTech Connect

    Zhang, Ke; Blake, Geoffrey A.; Bergin, Edwin A.

    2015-06-10

    Water and simple organic molecular ices dominate the mass of solid materials available for planetesimal and planet formation beyond the water snow line. Here we analyze ALMA long baseline 2.9, 1.3 and 0.87 mm continuum images of the young star HL Tau, and suggest that the emission dips observed are due to rapid pebble growth around the condensation fronts of abundant volatile species. Specifically, we show that the prominent innermost dip at 13 AU is spatially resolved in the 0.87 mm image, and its center radius is coincident with the expected mid-plane condensation front of water ice. In addition, two other prominent dips, at distances of 32 and 63 AU, cover the mid-plane condensation fronts of pure ammonia or ammonia hydrates and clathrate hydrates (especially with CO and N{sub 2}) formed from amorphous water ice. The spectral index map of HL Tau between 1.3 and 0.87 mm shows that the flux ratios inside the dips are statistically larger than those of nearby regions in the disk. This variation can be explained by a model with two dust populations, where most of the solid mass resides in a component that has grown to decimeter size scales inside the dips. Such growth is in accord with recent numerical simulations of volatile condensation, dust coagulation, and settling.

  3. The intracellular domain of Dumbfounded affects myoblast fusion efficiency and interacts with Rolling pebbles and Loner.

    PubMed

    Bulchand, Sarada; Menon, Sree Devi; George, Simi Elizabeth; Chia, William

    2010-02-23

    Drosophila body wall muscles are multinucleated syncytia formed by successive fusions between a founder myoblast and several fusion competent myoblasts. Initial fusion gives rise to a bi/trinucleate precursor followed by more fusion cycles forming a mature muscle. This process requires the functions of various molecules including the transmembrane myoblast attractants Dumbfounded (Duf) and its paralogue Roughest (Rst), a scaffold protein Rolling pebbles (Rols) and a guanine nucleotide exchange factor Loner. Fusion completely fails in a duf, rst mutant, and is blocked at the bi/trinucleate stage in rols and loner single mutants. We analysed the transmembrane and intracellular domains of Duf, by mutating conserved putative signaling sites and serially deleting the intracellular domain. These were tested for their ability to translocate and interact with Rols and Loner and to rescue the fusion defect in duf, rst mutant embryos. Studying combinations of double mutants, further tested the function of Rols, Loner and other fusion molecules. Here we show that serial truncations of the Duf intracellular domain successively compromise its function to translocate and interact with Rols and Loner in addition to affecting myoblast fusion efficiency in embryos. Putative phosphorylation sites function additively while the extreme C terminus including a PDZ binding domain is dispensable for its function. We also show that fusion is completely blocked in a rols, loner double mutant and is compromised in other double mutants. These results suggest an additive function of the intracellular domain of Duf and an early function of Rols and Loner which is independent of Duf.

  4. Evidence of Fast Pebble Growth Near Condensation Fronts in the HL Tau Protoplanetary Disk

    NASA Astrophysics Data System (ADS)

    Zhang, Ke; Blake, Geoffrey A.; Bergin, Edwin A.

    2015-06-01

    Water and simple organic molecular ices dominate the mass of solid materials available for planetesimal and planet formation beyond the water snow line. Here we analyze ALMA long baseline 2.9, 1.3 and 0.87 mm continuum images of the young star HL Tau, and suggest that the emission dips observed are due to rapid pebble growth around the condensation fronts of abundant volatile species. Specifically, we show that the prominent innermost dip at 13 AU is spatially resolved in the 0.87 mm image, and its center radius is coincident with the expected mid-plane condensation front of water ice. In addition, two other prominent dips, at distances of 32 and 63 AU, cover the mid-plane condensation fronts of pure ammonia or ammonia hydrates and clathrate hydrates (especially with CO and N2) formed from amorphous water ice. The spectral index map of HL Tau between 1.3 and 0.87 mm shows that the flux ratios inside the dips are statistically larger than those of nearby regions in the disk. This variation can be explained by a model with two dust populations, where most of the solid mass resides in a component that has grown to decimeter size scales inside the dips. Such growth is in accord with recent numerical simulations of volatile condensation, dust coagulation, and settling.

  5. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  6. Air Ingress Analyses on a High Temperature Gas-Cooled Reactor

    SciTech Connect

    Oh, Chang H; Moore, Richard Leroy; Merrill, Brad Johnson; Petti, David Andrew

    2001-11-01

    A primary-pipe break accident is one of the design-basis accidents of a high-temperature gas-cooled reactor (HTGR). When this accident occurs, air is anticipated to enter the reactor core from the break and oxidize the in-core graphite structure in the modular pebble bed reactor (MPBR). This paper presents the results of the graphite oxidation model developed as part of the Idaho National Engineering and Environmental Laboratory's Direct Research and Development effort. Although gas reactors have been tried in the past with limited success, the innovations of modularity and integrated state-ofart control systems coupled with improved fuel design and a pebble bed core make this design potentially very attractive from an economic and technical perspective. A schematic diagram on a reference design of the MPBR has been established on a major component level (INEEL & MIT, 1999). Steady-state and transient thermal hydraulics models will be produced with key parameters established for these conditions at all major components. Development of an integrated plant model to allow for transient analysis on a more sophisticated level is now being developed. In this paper, preliminary results of the hypothetical air ingress are presented. A graphite oxidation model was developed to determine temperature and the control mechanism in the spherical graphite geometry.

  7. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  8. Revision of the Western Australian pebble-mimic dragon species-group (Tympanocryptis cephalus: Reptilia: Agamidae).

    PubMed

    Doughty, Paul; Kealley, Luke; Shoo, Luke P; Melville, Jane

    2015-01-01

    Recent work on species complexes of the pebble-mimic dragons of the Australian genus Tympanocryptis has greatly clarified evolutionary relationships among taxa and also indicated that species diversity has been severely underestimated. Here we provide a morphological and molecular appraisal of variation in the T. cephalus species-group and find evidence for recognizing five species-level lineages from Western Australia. Four species-level lineages are strongly supported with a combined mitochondrial and nuclear DNA Bayesian analysis (a fifth population from the Gascoyne region lacked tissue samples). Morphologically, we found subtle, yet consistent, differences among the populations in scalation, color and pattern. True T. cephalus Günther is restricted to the coastal Pilbara region and characterized by five dark blotches on the dorsum, keeled ventrals, and other characters. Two other lineages within the Pilbara, from the Hamersley range and Fortescue/northern Pilbara region, differed from T. cephalus senso stricto by possessing a more elongate body and a plain dorsum. Furthermore, the Hamersley lineage differed from the Fortescue lineage by possessing slightly more reddish coloration and feeble keeling on the snout. Although there are few specimens and no tissue samples available for the Gascoyne population, these individuals are larger, have rugose scales on the snout, and possess scattered enlarged tubercles with three large blotches on the dorsum. The name T. cephalus gigas Mitchell is available for this population. The most widespread lineage, and the one best represented in collections and in field guides, occurs throughout central Western Australia. These Goldfield populations are characterized by a protruding snout, narrow rostral, and uniform reddish-brown coloration, often with a dark wash. Based on the genetic and morphological differences, we redescribe T. cephalus, resurrect and elevate T. gigas to a full species and designate a neotype for this taxon

  9. Revision of the Western Australian pebble-mimic dragon species-group (Tympanocryptis cephalus: Reptilia: Agamidae).

    PubMed

    Doughty, Paul; Kealley, Luke; Shoo, Luke P; Melville, Jane

    2015-11-03

    Recent work on species complexes of the pebble-mimic dragons of the Australian genus Tympanocryptis has greatly clarified evolutionary relationships among taxa and also indicated that species diversity has been severely underestimated. Here we provide a morphological and molecular appraisal of variation in the T. cephalus species-group and find evidence for recognizing five species-level lineages from Western Australia. Four species-level lineages are strongly supported with a combined mitochondrial and nuclear DNA Bayesian analysis (a fifth population from the Gascoyne region lacked tissue samples). Morphologically, we found subtle, yet consistent, differences among the populations in scalation, color and pattern. True T. cephalus Günther is restricted to the coastal Pilbara region and characterized by five dark blotches on the dorsum, keeled ventrals, and other characters. Two other lineages within the Pilbara, from the Hamersley range and Fortescue/northern Pilbara region, differed from T. cephalus senso stricto by possessing a more elongate body and a plain dorsum. Furthermore, the Hamersley lineage differed from the Fortescue lineage by possessing slightly more reddish coloration and feeble keeling on the snout. Although there are few specimens and no tissue samples available for the Gascoyne population, these individuals are larger, have rugose scales on the snout, and possess scattered enlarged tubercles with three large blotches on the dorsum. The name T. cephalus gigas Mitchell is available for this population. The most widespread lineage, and the one best represented in collections and in field guides, occurs throughout central Western Australia. These Goldfield populations are characterized by a protruding snout, narrow rostral, and uniform reddish-brown coloration, often with a dark wash. Based on the genetic and morphological differences, we redescribe T. cephalus, resurrect and elevate T. gigas to a full species and designate a neotype for this taxon

  10. An exploration hydrogeochemical study at the giant Pebble porphyry Cu-Au-Mo deposit, Alaska, USA, using high-resolution ICP-MS

    USGS Publications Warehouse

    Eppinger, Robert G.; Fey, David L.; Giles, Stuart A.; Kelley, Karen D.; Smith, Steven M.

    2012-01-01

    A hydrogeochemical study using high resolution ICP-MS was undertaken at the giant Pebble porphyry Cu-Au-Mo deposit and surrounding mineral occurrences. Surface water and groundwater samples from regional background and the deposit area were collected at 168 sites. Rigorous quality control reveals impressive results at low nanogram per litre (ng/l) levels. Sites with pH values below 5.1 are from ponds in the Pebble West area, where sulphide-bearing rubble crop is thinly covered. Relative to other study area waters, anomalous concentrations of Cu, Cd, K, Ni, Re, the REE, Tl, SO42− and F− are present in water samples from Pebble West. Samples from circum-neutral waters at Pebble East and parts of Pebble West, where cover is much thicker, have anomalous concentrations of Ag, As, In, Mn, Mo, Sb, Th, U, V, and W. Low-level anomalous concentrations for most of these elements were also found in waters surrounding nearby porphyry and skarn mineral occurrences. Many of these elements are present in low ng/l concentration ranges and would not have been detected using traditional quadrupole ICP-MS. Hydrogeochemical exploration paired with high resolution ICP-MS is a powerful new tool in the search for concealed deposits.

  11. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    SciTech Connect

    DeHart, Mark D; Pritchard, Megan L

    2008-01-01

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the results of

  12. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  13. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  14. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  15. Geological analysis of aeromagnetic data from southwestern Alaska: implications for exploration in the area of the Pebble porphyry Cu-Au-Mo deposit

    USGS Publications Warehouse

    Anderson, Eric D.; Hitzman, Murray W.; Monecke, Thomas; Bedrosian, Paul A.; Shah, Anjana K.; Kelley, Karen D.

    2013-01-01

    Aeromagnetic data are used to better understand the geology and mineral resources near the Late Cretaceous Pebble porphyry Cu-Au-Mo deposit in southwestern Alaska. The reduced-to-pole (RTP) transformation of regional-scale aeromagnetic data shows that the Pebble deposit is within a cluster of magnetic anomaly highs. Similar to Pebble, the Iliamna, Kijik, and Neacola porphyry copper occurrences are in magnetic highs that trend northeast along the crustal-scale Lake Clark fault. A high-amplitude, short- to moderate-wavelength anomaly is centered over the Kemuk occurrence, an Alaska-type ultramafic complex. Similar anomalies are found west and north of Kemuk. A moderate-amplitude, moderate-wavelength magnetic low surrounded by a moderate-amplitude, short-wavelength magnetic high is associated with the gold-bearing Shotgun intrusive complex. The RTP transformation of the district-scale aeromagnetic data acquired over Pebble permits differentiation of a variety of Jurassic to Tertiary magmatic rock suites. Jurassic-Cretaceous basalt and gabbro units and Late Cretaceous biotite pyroxenite and granodiorite rocks produce magnetic highs. Tertiary basalt units also produce magnetic highs, but appear to be volumetrically minor. Eocene monzonite units have associated magnetic lows. The RTP data do not suggest a magnetite-rich hydrothermal system at the Pebble deposit. The 10-km upward continuation transformation of the regional-scale data shows a linear northeast trend of magnetic anomaly highs. These anomalies are spatially correlated with Late Cretaceous igneous rocks and in the Pebble district are centered over the granodiorite rocks genetically related to porphyry copper systems. The spacing of these anomalies is similar to patterns shown by the numerous porphyry copper deposits in northern Chile. These anomalies are interpreted to reflect a Late Cretaceous magmatic arc that is favorable for additional discoveries of Late Cretaceous porphyry copper systems in southwestern

  16. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  17. The aluminum phosphate zone in the Peace River area, land-pebble phosphate field, Florida

    USGS Publications Warehouse

    Cathcart, James B.

    1953-01-01

    The Peace River area, comprising T. 30 and 31 S., R. 24 and 25 E., contains a thicker and more persistent aluminum phosphate zone, and one that is higher in P2O5 and uranium content than is known elsewhere in the land-pebble phosphate district. This report has been prepared to bring together all of the information on the aluminum phosphate zone in the area where the first plant to treat this material will probably be located. The area may be divided into three physiographic units, (1) the ridge, (2) the flatwoods, and (3) the valley. Maps showing distribution and grade of the aluminum phosphate zone indicate that the zone is thin or absent in the ridge unit, thickest and most persistent, and of the best grade in P2O5 and uranium in the flatwoods unit, and absent or very low in grade in the valley unit. Maps of thickness and of chemical composition show that even in favorable areas there are places where the aluminum phosphate zone is missing or of questionable economic importance. The distribution maps also show that areas of high P2O5 and high uranium content coincide closely. Areas containing thick aluminum phosphate material usually have high uranium and P2O5 contents. It is estimated that an average of 13,000 tons per day of aluminum phosphate material might be mined from this area. This figure is based on the probable amount of time, per year, that mining would be in favorable ground. When all mines in the area are in favorable ground, the tonnage per day might be about 23,000 tons. Tonnages of aluminum phosphate material have been computed for about 36 percent of the area of T. 30 S., R. 25 E., and for 18 percent of the area of T. 31 S., R. 25 E. The total inferred tonnage is about 150,000,000 short tons, with an average grade of 0.012 percent U3O8.

  18. NUCLEAR REACTOR

    DOEpatents

    Moore, R.V.; Bowen, J.H.; Dent, K.H.

    1958-12-01

    A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

  19. A pebbles accretion model with chemistry and implications for the solar system in the lights of Juno

    NASA Astrophysics Data System (ADS)

    Ali-Dib, Mohamad

    2016-10-01

    The chemical compositions of the solar system giant planets are a major source of informations on their origins. Since the measurements by the Galileo probe, multiple models have been put forward to try and explain the noble gases enrichment in Jupiter. The most discussed among these are its formation in the outer cold nebula and its formation in a partially photoevaporated disk. In this work I couple a pebbles accretion model to the disk's chemistry and photoevaporation in order to make predictions from both scenarios and compare them to the upcoming Juno measurements. The model include pebbles and gas accretion, type I and II migration, photoevaporation and chemical measurements from meteorites, comets and disks. Population synthesis simulations are used to explore the models free parameters (planets initial conditions), where then the results are narrowed down using the planets chemical, dynamical and core mass costraints. We end up with a population that fits all of the constrains. These are then used to predict the oxygen abundance and core mass in Jupiter, to be compared to results of Juno. Same calculations are also done for Saturn and Neptune for comparison. I will present the results from these simulations as well as the predictions from all of the different models.Ali-Dib, M. (2016ab, submitted to MNRAS)

  20. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  1. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  2. Chemical Reactors.

    ERIC Educational Resources Information Center

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  3. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  4. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  5. REACTOR SHIELD

    DOEpatents

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  6. NUCLEAR REACTOR

    DOEpatents

    Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

    1962-10-23

    A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

  7. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  8. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  9. NUCLEAR REACTOR

    DOEpatents

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  10. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  11. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  12. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    SciTech Connect

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  13. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  14. A novel reusable nanocomposite for complete removal of dyes, heavy metals and microbial load from water based on nanocellulose and silver nano-embedded pebbles.

    PubMed

    Suman; Kardam, Abhishek; Gera, Meeta; Jain, V K

    2015-01-01

    The present work proposed a nanocellulose (NC)-silver nanoparticles (AgNPs) embedded pebbles-based composite material as a novel reusable cost-effective water purification device for complete removal of dyes, heavy metals and microbes. NC was prepared using acid hydrolysis of cellulose. The AgNPs were generated in situ using glucose and embedded within the porous concrete pebbles by the technique of inter-diffusion of ion, providing a very strong binding of nanoparticles within the porous pebbles and thus preventing any nanomaterials leaching. Fabrication of a continual running water purifier was achieved by making different layering of NC and Ag nano-embedded pebbles in a glass column. The water purifier exhibited not only excellent dye and heavy metal adsorption capacity, but also long-term antibacterial activity against pathogenic and non-pathogenic bacterial strains. The adsorption mainly occurred through electrostatic interaction and pore diffusion also contributed to the process. The bed column purifier has shown 99.48% Pb(II) and 98.30% Cr(III) removal efficiency along with 99% decontamination of microbial load at an optimum working pH of 6.0. The high adsorption capacity and reusability, with complete removal of dyes, heavy metals and Escherichia coli from the simulated contaminated water of composite material, will provide new opportunities to develop a cost-effective and eco-friendly water purifier for commercial application. PMID:25243917

  15. A novel reusable nanocomposite for complete removal of dyes, heavy metals and microbial load from water based on nanocellulose and silver nano-embedded pebbles.

    PubMed

    Suman; Kardam, Abhishek; Gera, Meeta; Jain, V K

    2015-01-01

    The present work proposed a nanocellulose (NC)-silver nanoparticles (AgNPs) embedded pebbles-based composite material as a novel reusable cost-effective water purification device for complete removal of dyes, heavy metals and microbes. NC was prepared using acid hydrolysis of cellulose. The AgNPs were generated in situ using glucose and embedded within the porous concrete pebbles by the technique of inter-diffusion of ion, providing a very strong binding of nanoparticles within the porous pebbles and thus preventing any nanomaterials leaching. Fabrication of a continual running water purifier was achieved by making different layering of NC and Ag nano-embedded pebbles in a glass column. The water purifier exhibited not only excellent dye and heavy metal adsorption capacity, but also long-term antibacterial activity against pathogenic and non-pathogenic bacterial strains. The adsorption mainly occurred through electrostatic interaction and pore diffusion also contributed to the process. The bed column purifier has shown 99.48% Pb(II) and 98.30% Cr(III) removal efficiency along with 99% decontamination of microbial load at an optimum working pH of 6.0. The high adsorption capacity and reusability, with complete removal of dyes, heavy metals and Escherichia coli from the simulated contaminated water of composite material, will provide new opportunities to develop a cost-effective and eco-friendly water purifier for commercial application.

  16. NEUTRONIC REACTOR

    DOEpatents

    Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

    1958-09-01

    This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

  17. Porphyry Cu indicator minerals in till as an exploration tool: Example from the giant pebble porphyry Cu-Au-Mo deposit, Alaska, USA

    USGS Publications Warehouse

    Kelley, Karen D.; Eppinger, Robert G.; Lang, J.; Smith, Steven M.; Fey, David L.

    2011-01-01

    Porphyry Cu indicator minerals are mineral species in clastic sediments that indicate the presence of mineralization and hydrothermal alteration associated with porphyry Cu and associated skarn deposits. Porphyry Cu indicator minerals recovered from shallow till samples near the giant Pebble Cu-Au-Mo porphyry deposit in SW Alaska, USA, include apatite, andradite garnet, Mn-epidote, visible gold, jarosite, pyrite, and cinnabar. Sulphide minerals other than pyrite are absent from till, most likely due to the oxidation of the till. The distribution of till samples with abundant apatite and cinnabar suggest sources other than the Pebble deposit. With three exceptions, all till samples up-ice of the Pebble deposit contain 40grains/10kg) are in close proximity to smaller porphyry and skarn occurrences in the region. The distribution of Mn-epidote closely mimics the distribution of garnet in the till samples and further supports the interpretation that these minerals most likely reflect skarns associated with the porphyry deposits. All but two till samples, including those up-ice from the deposit, contain some gold grains. However, tills immediately west and down-ice of Pebble contain more abundant gold grains, and the overall number of grains decreases in the down-ice direction. Furthermore, all samples in the immediate vicinity of Pebble contain more than 65% pristine and modified grains compared to mostly re-shaped grains in distal samples. The pristine gold in till reflects short transport distances and/or liberation of gold during in-situ weathering of transported chalcopyrite grains. Jarosite is also abundant (1-2 500 grains/10kg) in samples adjacent to and up to 7 km down-ice from the deposit. Most jarosite grains are rounded and preliminary Ar/Ar dates suggest the jarosite formed prior to glaciation and it implies that a supergene cap existed over Pebble West. Assuming this interpretation is accurate, it suggests a shallow level of erosion of the Pebble deposit by

  18. Exploration case study using indicator minerals in till at the giant Pebble porphyry Cu-Au-Mo deposit, southwest Alaska, USA

    USGS Publications Warehouse

    Eppinger, Robert G.; Kelley, Karen D.; Fey, David L.; Giles, Stuart A.; Smith, Steven G.

    2011-01-01

    The Pebble deposit in southwest Alaska (Fig. 1) contains one of the largest resources of copper and gold in the world. It includes a measured and indicated resource of 5,942 million tonnes (Mt) at 0.42% Cu, 0.35 g/t Au, and 250 ppm Mo (0.30% copper equivalent, CuEQ, cut off) and contains significant concentrations of Ag, Pd, and Re (Northern Dynasty Minerals 2011). The deposit remains open at depth. The Pebble West zone was discovered in 1989 by Cominco American. In 2005, Northern Dynasty Minerals Ltd. (NDM) discovered Pebble East, and in July 2007, NDM partnered with Anglo American to form the Pebble Limited Partnership (PLP). The U.S. Geological Survey began collaborative investigations with PLP in 2007 to identify techniques that will improve mineral exploration in covered terranes. The Pebble deposit is an ideal location for such a study because the deposit is undisturbed (except for drilling), is almost entirely concealed by post-mineral volcanic rocks and glacial deposits, and because its distribution is well constrained in the subsurface by PLP’s drill-hole geology and geochemistry. An exploration method developed by Averill (2007) that utilizes porphyry copper indicator minerals (PCIMR) in glacial till samples was applied at Pebble; samples were collected up- and down-ice (of former glaciers) from the deposit. The distribution of several PCIMs identifies the deposit, which suggests that PCIMs may be useful in exploration for other concealed porphyry deposits in the region. In this study, we compare the efficacy of PCIMs relative to that of pond and stream sediments also collected in the deposit area. The Pebble deposit is located 380 km southwest of Anchorage, in the Bristol Bay region of southwest Alaska. There is no road network and access to the study area is by helicopter. The deposit is situated in a broad glacially sculpted topographic low at the head of three drainages, Talarik Creek, North Fork Koktuli River, and the South Fork Koktuli River (Fig

  19. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

  20. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.

    1957-10-01

    A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

  1. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  2. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  3. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

    A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

  4. Catalytic reactor

    SciTech Connect

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  5. NEUTRONIC REACTORS

    DOEpatents

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  6. Three-dimensional distribution of igneous rocks near the Pebble porphyry Cu-Au-Mo deposit in southwestern Alaska: constraints from regional-scale aeromagnetic data

    USGS Publications Warehouse

    Anderson, Eric D.; Zhou, Wei; Li, Yaoguo; Hitzman, Murray W.; Monecke, Thomas; Lang, James R.; Kelley, Karen D.

    2014-01-01

    Aeromagnetic data helped us to understand the 3D distribution of plutonic rocks near the Pebble porphyry copper deposit in southwestern Alaska, USA. Magnetic susceptibility measurements showed that rocks in the Pebble district are more magnetic than rocks of comparable compositions in the Pike Creek–Stuyahok Hills volcano-plutonic complex. The reduced-to-pole transformation of the aeromagnetic data demonstrated that the older rocks in the Pebble district produce strong magnetic anomaly highs. The tilt derivative transformation highlighted northeast-trending lineaments attributed to Tertiary volcanic rocks. Multiscale edge detection delineated near-surface magnetic sources that are mostly outward dipping and coalesce at depth in the Pebble district. The total horizontal gradient of the 10-km upward-continued magnetic data showed an oval, deep magnetic contact along which porphyry deposits occur. Forward and inverse magnetic modeling showed that the magnetic rocks in the Pebble district extend to depths greater than 9 km. Magnetic inversion was constrained by a near-surface, 3D geologic model that is attributed with measured magnetic susceptibilities from various rock types in the region. The inversion results indicated that several near-surface magnetic sources with moderate susceptibilities converge with depth into magnetic bodies with higher susceptibilities. This deep magnetic source appeared to rise toward the surface in several areas. An isosurface value of 0.02 SI was used to depict the magnetic contact between outcropping granodiorite and nonmagnetic sedimentary host rocks. The contact was shown to be outward dipping. At depths around 5 km, nearly the entire model exceeded the isosurface value indicating the limits of nonmagnetic host material. The inversion results showed the presence of a relatively deep, northeast-trending magnetic low that parallels lineaments mapped by the tilt derivative. This deep low represents a strand of the Lake Clark fault.

  7. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

    This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

  8. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  9. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  10. A Hydrodynamic Investigation of Instream Pebble Clusters in Gravel-Bed Rivers With Implications for Fish Habitat

    NASA Astrophysics Data System (ADS)

    Lacey, R. J.; Roy, A. G.

    2006-05-01

    There does not appear to be a straight forward relationship between fish behaviour and in-stream large roughness elements such as boulders and pebble clusters. Studies have found increased salmonid densities following the placement of boulder clusters attributed to increased habitat complexity - variability in depth, cover, and current velocity. Yet laboratory bioenergetics studies have indicated that the complex environment induced by added boulders and cobbles reduces drift feeding and increases energy expenditure in juvenile salmon. These somewhat conflicting results may be related to an inadequate representation of the hydrodynamics associated with the large roughness elements in the studies. Most ecological fish habitat studies do not include a full or even a partial characterization of the turbulent flow dynamics even though positive relationships have been found between the swimming costs of salmonids and flow turbulence. This study provides a spatial description of the turbulent flow field associated with a naturally formed instream pebble cluster with a relative roughness of 0.3. The mean streamwise velocity and water depth at the time of measurement were 73 cm/s, and 35 cm, respectively. The hydrodynamic characterization presented here may aid in understanding the mechanisms involved in the linkage between large roughness elements and salmon productivity. Instantaneous high frequency velocities were measured using the simultaneous deployment of four acoustic Doppler velocimeters. Streamwise component mean velocities are reduced to near zero values in the near wake of the pebble cluster while turbulent kinetic energy values increase by a factor of approximately 2.0 compared with velocities outside the wake. The vertical velocity spectra estimated from measurements within the wake zone contain a marked peak at approximately 1.5 Hz associated with eddies shedding from the shear layer in the lee of the cluster. Turbulent events detected using the U

  11. Characteristics of Pebble- and Cobble-Sized Clasts Along the Curiosity Rover Traverse from Sol 0 to 800.

    NASA Astrophysics Data System (ADS)

    Aileen Yingst, R.; Cropper, Kevin; Blank, Jennifer; Goetz, Walter; Hamilton, Victoria; Hipkin, Victoria; Kah, Linda; Madsen, Morten Bo; Newsom, Horton; Williams, Rebecca; Bridges, John; Martinez-Frias, Jesús; King, Penelope

    2015-04-01

    The transport, sorting and abrasive processes that modify the loose surface fragments comprising a sedimentary population are best recorded in the morphologic characteristics of those fragments (also termed clasts or particles). Here we assess morphologic characteristics (size, shape, roundness, texture) of clasts in the pebble to cobble size range (2-256 mm) in Gale Crater imaged along the path of the Curisoty rover from sols 0 to 800. Pebble- to cobble-sized clasts along Curiosity's traverse most likely include geologic materials from the walls of fluvial canyons that debouch onto the crater floor, fragments shed from the central mound of the crater, and grains from modern eolian dunes. Our goal is to help constrain the boundaries of potential transport mechanisms important throughout the geologic history of the crater interior. Clast survey observations were taken on 162 sols. Most common clast types include: Type 1. These grey, fine-grained clasts are the most common type between sols 0 to ~650. Particles are angular to sub-angular (though sub-rounded clasts appear beginning ~sol 548), with flat facets terminating in sharp or slightly rounded edges. Surface texture is often smooth, but fresher facets can be rough and knobby at the sub-mm scale; occasionally, faint layers can be discerned. Wind-eroded features are common. Type 2. These clasts are gray and angular to sub-angular, displaying faces with circular or elongated concave-outward ellipsoids averaging 0.5-1.5 mm long-axis. A potential variant of this clast type is one in which ellipsoids are so deeply weathered that the particle takes on a scoria-like shape. Type 3. This class consists of angular to sub-angular void-rich clasts. These are rare, and disappear after ~sol 50, but reappear around sol 672. Type 4. These particles are angular to sub-angular and clast-rich. Clasts protrude as they wear, making surface textures jagged-looking. Type 5. Particles in this class vary in color, are equidimensional or

  12. Preliminary notes on distribution of uranium in the Florida pebble phosphate field and suggestions for studying and sampling

    USGS Publications Warehouse

    ,

    1948-01-01

    The accompanying map and sections show examples of the present state of information about the occurrence of the "Leached" uranium-bearing bed in the Florida pebble phosphate district. The dashed lines on the map define, as closely as present data permit, the limit of the area in which this bed contains significant amounts of uranium. The figures next to localities on the map indicate first, the thickness of the bed in feet; and second the uranium content in thousandths of percent. For example, the figures 16-10 next to the TVA localities in Secs. 9 and 10, T. 32 S., R. 26 E. indicate 16 feet at 0.010 percent uranium. A "0" by a locality indicates either that the uranium content is less than 0.001 percent or less than the concentration in the underlying phosphate beds (matrix of the miners) or that the leached bed is not present.

  13. First-order control of surface roughness at three scales: boundary layer dynamics, tracer dispersion and pebble abrasion

    NASA Astrophysics Data System (ADS)

    Jerolmack, D. J.; Litwin, K. L.; Phillips, C. B.; Martin, R. L.

    2012-12-01

    In many situations it may be appropriate to treat surfaces as smooth and particles as spherical, however here we focus on scenarios in which the roughness of the surface exerts a first-order control on flow and transport dynamics. We describe three vignettes at three different scales: (1) roughness transitions and resulting sediment transport dynamics over ~10-km distance in a desert dune field; (2) reach-scale river bed roughness and its influence on dispersion of tracer particles in bed load; and (3) the control of particle surface roughness on the nature and rate of pebble abrasion. For (1), we show how the abrupt transition from a flat surface to a dune field may be treated as a step increase in the aerodynamic roughness parameter - so long as the spatial scale considered is significantly larger than that of an individual dune. This increase causes a spatial decline in the boundary stress downwind that may be understood using simple boundary layer theory, resulting in a factor of three decrease in the sand flux over a distance of kilometers. For (2), laboratory and field studies of tracer particles in bed load indicate that they undergo short flights separated by long rest periods having a power-law tail - even in steady flows. We hypothesize that for near-threshold transport - which predominates is coarse-grained rivers - particles become trapped in 'wells' produced by surface roughness, and their rest time is controlled by the time for the surface to scour down and release them. Laboratory observations support this hypothesis, while comparison to non-geophysical 'flows' indicates that these dynamics are generic to transport in disordered systems. Finally, for (3) we report laboratory experiments by our group and others showing how abrasion rate decreases with decreasing particle roughness. Geometric models quantitatively support the intuition that locations of high positive curvature on pebble surfaces are more susceptible to abrasion; as they are

  14. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  15. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  16. NEUTRONIC REACTOR

    DOEpatents

    Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

    1959-10-27

    BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

  17. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  18. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1962-12-25

    A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

  19. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

    The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

  20. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation. PMID:27573503

  1. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1960-09-27

    A unit assembly is described for a neutronic reactor comprising a tube and plurality of spaced parallel sandwiches in the tube extending lengthwise thereof, each sandwich including a middle plate having a central opening for plutonium and other openings for fertile material at opposite ends of the plate.

  2. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  3. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

    A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

  4. REACTOR MONITORING

    DOEpatents

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  5. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  6. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  7. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

    This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

  8. 3-D transient analysis of pebble-bed HTGR by TORT-TD/ATTICA3D

    SciTech Connect

    Seubert, A.; Sureda, A.; Lapins, J.; Buck, M.; Bader, J.; Laurien, E.

    2012-07-01

    As most of the acceptance criteria are local core parameters, application of transient 3-D fine mesh neutron transport and thermal hydraulics coupled codes is mandatory for best estimate evaluations of safety margins. This also applies to high-temperature gas cooled reactors (HTGR). Application of 3-D fine-mesh transient transport codes using few energy groups coupled with 3-D thermal hydraulics codes becomes feasible in view of increasing computing power. This paper describes the discrete ordinates based coupled code system TORT-TD/ATTICA3D that has recently been extended by a fine-mesh diffusion solver. Based on transient analyses for the PBMR-400 design, the transport/diffusion capabilities are demonstrated and 3-D local flux and power redistribution effects during a partial control rod withdrawal are shown. (authors)

  9. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

    A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

  10. NUCLEAR REACTORS

    DOEpatents

    Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

    1961-12-01

    An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

  11. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  12. NEUTRONIC REACTORS

    DOEpatents

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  13. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  14. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

    This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

  15. Optimizing Neutron Thermal Scattering Effects in very High Temperature Reactors. Final Report

    SciTech Connect

    Hawari, Ayman; Ougouag, Abderrafi

    2014-07-08

    This project aims to develop a holistic understanding of the phenomenon of neutron thermalization in the VHTR. Neutron thermalization is dependent on the type and structure of the moderating material. The fact that the moderator (and reflector) in the VHTR is a solid material will introduce new and interesting considerations that do not apply in other (e.g. light water) reactors. The moderator structure is expected to undergo radiation induced changes as the irradiation (or burnup) history progresses. In this case, the induced changes in structure will have a direct impact on many properties including the neutronic behavior. This can be easily anticipated if one recognizes the dependence of neutron thermalization on the scattering law of the moderator. For the pebble bed reactor, it is anticipated that the moderating behavior can be tailored, e.g. using moderators that consist of composite materials, which could allow improved optimization of the moderator-to-fuel ratio.

  16. Effect on the tritium breeding ratio for a distributed ICRF antenna in a DEMO reactor

    NASA Astrophysics Data System (ADS)

    Garcia, A.; Noterdaeme, J.-M.; Fischer, U.; Dies, J.

    2015-12-01

    The paper reports results of MCNP-5 calculations to assess the effect on the Tritium Breeding Ratio (TBR) when integrating a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna in the blanket of DEMO fusion power reactor. The calculations consider different parameters such as the ICRF covering ratio and the type of breeding blanket including the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. For an antenna with a full toroidal circumference of 360°, located poloidally at 40° with a poloidal extension of 1 m, the reduction of the TBR is -0.349% for the HCPB blanket and -0.532% for the HCLL blanket. The distributed ICRF antenna is thus shown to have only a marginal effect on the TBR of the DEMO reactor.

  17. Effect on the tritium breeding ratio for a distributed ICRF antenna in a DEMO reactor

    SciTech Connect

    Garcia, A.; Noterdaeme, J.-M.; Fischer, U.; Dies, J.

    2015-12-10

    The paper reports results of MCNP-5 calculations to assess the effect on the Tritium Breeding Ratio (TBR) when integrating a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna in the blanket of DEMO fusion power reactor. The calculations consider different parameters such as the ICRF covering ratio and the type of breeding blanket including the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) concepts. For an antenna with a full toroidal circumference of 360°, located poloidally at 40° with a poloidal extension of 1 m, the reduction of the TBR is −0.349% for the HCPB blanket and −0.532% for the HCLL blanket. The distributed ICRF antenna is thus shown to have only a marginal effect on the TBR of the DEMO reactor.

  18. Geophysical and geochemical data from the area of the Pebble Cu-Au-Mo porphyry deposit, southwestern Alaska: Contributions to assessment techniques for concealed mineral resources

    USGS Publications Warehouse

    Anderson, E.D.; Smith, S.M.; Giles, S.A.; Granitto, Matthew; Eppinger, R.G.; Bedrosian, P.A.; Shah, A.K.; Kelley, K.D.; Fey, D.L.; Minsley, B.J.; Brown, P.J.

    2011-01-01

    In 2007, the U.S. Geological Survey began a multidisciplinary study in southwest Alaska to investigate the setting and detectability of mineral deposits in concealed volcanic and glacial terranes. The study area hosts the world-class Pebble porphyry Cu-Au-Mo deposit, and through collaboration with the Pebble Limited Partnership, a range of geophysical and geochemical investigations was carried out in proximity to the deposit. The deposit is almost entirely concealed by tundra, glacial deposits, and post-mineralization volcanic rocks. The discovery of mineral resources beneath cover is becoming more important because most of the mineral resources at the surface have already been discovered. Research is needed to identify ways in which to assess for concealed mineral resources. This report presents the uninterpreted geophysical measurements and geochemical and mineralogical analytical data from samples collected during the summer field seasons from 2007 to 2010, and makes the data available in a single Geographic Information System (GIS) database.

  19. Reactor applications of the Compact Fusion Advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Hoffman, H. A.; Logan, B. G.; Campbell, R. B.

    1988-03-01

    A preliminary design of a D-T fusion reactor blanket and MHD power conversion system is made based on the CFAR concept, and it was found that performance and costs for the reference cycle are very attractive. While much remains to be done, the potential advantage of liquid metal Rankine cycles for fusion applications are much clearer now. These include low pressures and mass flow rates, a nearly isothermal module shell which minimizes problems of thermal distortion and stresses, and an insensitivity to pressure losses in the blanket so that the two-phase MHD pressure drops in the boiling part of the blanket and the ordinary vapor pressure drops in the pebble-bed superheating zones are acceptable (the direct result of pumping a liquid rather than having to compress a gas). There are no moving parts in the high-temperature MHD power generators, no steam bottoming plant is required, only small vapor precoolers and condensers are needed because of the high heat rejection temperatures, and only a relatively small natural-draft heat exchanger is required to reject the heat to the atmosphere. The net result is a very compact fusion reactor and power conversion system which fit entirely inside an 18 meter radius reactor vault. Although a cost analysis has not yet been performed, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity.

  20. Contingency table analysis of pebble lithology and roundness: A case study of Huangshui River, China and comparison to rivers in the Rocky Mountains, USA

    USGS Publications Warehouse

    Miao, X.; Lindsey, D.A.; Lai, Z.; Liu, Xiuying

    2010-01-01

    Contingency table analysis of pebble lithology and roundness is an effective way to identify the source terrane of a drainage basin and to distinguish changes in basin size, piracy, tectonism, and other events. First, the analysis to terrace gravel deposited by the Huangshui River, northeastern Tibet Plateau, China, shows statistically contrasting pebble populations for the oldest terrace (T7, Dadongling, 1.2. Ma) and the youngest terraces (T0-T3, ?. 0.15. Ma). Two fluvial processes are considered to explain the contrast in correlation between lithology and roundness in T7 gravel versus T0-T3 gravel: 1) reworking of T7 gravel into T0-T3 gravel and 2) growth in the size of the river basin between T7 and T0-T3 times. We favor growth in basin size as the dominant process, from comparison of pebble counts and contingency tables. Second, comparison of results from Huangshui River of China to three piedmont streams of the Rocky Mountains, USA highlights major differences in source terrane and history. Like Rocky Mountain piedmont gravel from Colorado examples, the Huangshui gravels show a preference (observed versus expected frequency) for rounded granite. But unlike Rocky Mountain gravel, Huangshui gravel shows a preference for angular quartzite and for rounded sandstone. In conclusion, contrasting behavior of lithologies during transport, not always apparent in raw pebble counts, is readily analyzed using contingency tables to identify the provenance of individual lithologies, including recycled clasts. Results of the analysis may help unravel river history, including changes in basin size and lithology. ?? 2009.

  1. Pb-Sr-Nd isotopes in surficial materials at the Pebble Porphyry Cu-Au-Mo Deposit, Southwestern Alaska: can the mineralizing fingerprint be detected through cover?

    USGS Publications Warehouse

    Ayuso, Robert A.; Kelley, Karen D.; Eppinger, Robert G.; Forni, Francesca

    2013-01-01

    The Cretaceous Pebble porphyry Cu-Au-Mo deposit is covered by tundra and glacigenic sediments. Pb-Sr-Nd measurements were done on sediments and soils to establish baseline conditions prior to the onset of mining operations and contribute to the development of exploration methods for concealed base metal deposits of this type. Pebble rocks have a moderate range for 206Pb/204Pb = 18.574 to 18.874, 207Pb/204Pb = 15.484 to 15.526, and 208,Pb/204Pb = 38.053 to 38.266. Mineralized granodiorite shows a modest spread in 87Sr/86Sr (0.704354–0.707621) and 143Nd/144Nd (0.512639–0.512750). Age-corrected (89 Ma) values for the granodiorite yield relatively unradiogenic Pb (e.g., 207Pb/204Pb 87Sr/86Sr, and positive values of ɛNd (1.00–4.52) that attest to a major contribution of mantle-derived source rocks. Pond sediments and soils have similar Pb isotope signatures and 87Sr/86Sr and 143Nd/144Nd values that resemble the mineralized granodiorites. Glacial events have obscured the recognition of isotope signatures of mineralized rocks in the sediments and soils. Baseline radiogenic isotope compositions, prior to the onset of mining operations, reflect natural erosion, transport and deposition of heterogeneous till sheets that included debris from barren rocks, mineralized granodiorite and sulfides from the Pebble deposit, and other country rocks that pre- and postdate the mineralization events. Isotopic variations suggest that natural weathering of the deposit is generally reflected in these surficial materials. The isotope data provide geochemical constraints to glimpse through the extensive cover and together with other geochemical observations provide a vector to concealed mineralized rocks genetically linked with the Pebble deposit.

  2. Late Jurassic blueschist facies pebbles from the Western Carpathian orogenic wedge and paleostructural implications for Western Tethys evolution

    NASA Astrophysics Data System (ADS)

    Dal Piaz, Giorgio V.; Martin, Silvana; Villa, Igor M.; Gosso, Guido; Marschalko, Robert

    1995-08-01

    In spite of the absence of ophiolitic slices at the surface, some traces of the lost Tethys ocean are recorded along the Pieniny Klippen Belt (PKB), a narrow décollement thrust system sutured at the transpressive boundary between the Outer and Inner Carpathians. The enigmatic precollisional evolution of Western Carpathians can be deciphered from some late Albian to Campanian flysch conglomerates which display chrome spinel grains, ophiolitic detritus and pebbles of blueschist facies tholeiitic metabasalts yielding a 40Ar/39Ar plateau age of 155.4+/-0.6 Ma. Other detrital components are represented by extrabasinal pebbles of limestones, arc volcanics, and igneous to metamorphic basement rocks from southern sources. Our results suggest a markedly northward extension of the sublongitudinal Triassic Vardar (Meliata) Ocean and its subduction since the late Middle Jurassic, supposedly balanced westward by coeval spreading in the Ligurian-Piedmont basin of the Apennine-Western Alpine Tethys. A lateral kinematic connection between these diachronous and roughly parallel Tethys branches was provided on the north by a left-lateral east-west trending shear zone running from the Swiss-Austrian Penninic domain to the Northern Carpathians. This reconstruction replaces the classic model of two paired North Penninic and South Penninic oceanic basins and eastern homologues with the Briançonnais-Hochstegen and Czorstin microcontinents in between. The Late Jurassic-Early Cretaceous evolution of the Carpathian active margin was characterized by subduction metamorphism and accretion of a wide orogenic wedge; in this time, the shallowing to deeply subsiding basins inferred from facies analyses on the sedimentary units of the PKB were likely floored by individual sections of the growing wedge. Later, some exhuming blueschist ophiolitic units of the wedge were uplifted to the surface and functioned in the Albian-Campanian as an ``exotic ridge'' supplying clasts to the forearc basin

  3. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    SciTech Connect

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  4. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    SciTech Connect

    Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns; Fugate, David L; Holcomb, David Eugene; Kisner, Roger A; Peretz, Fred J; Robb, Kevin R; Wilgen, John B; Wilson, Dane F

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  5. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  6. Age and provenance of mica-schist pebbles from the Eocene conglomerates of the Tylicz and Krynica Zone (Magura Nappe, Outer Flysch Carpathians)

    NASA Astrophysics Data System (ADS)

    Oszczypko, Nestor; Salata, Dorota; Konečný, Patrik

    2016-06-01

    During the Łate Cretaceous to Palaeogene, the Magura Basin was supplied by clastic material from source areas situated on the northern and southern margins of the basin, which do not outcrop on the surface at present. The northern source area is traditionally connected with the Silesian Ridge, whereas the position of the southern one is still under discussion. A source area situated SE of the Magura Basin supplied the Eocene pebbly para-conglomerates containing partly exotic material. The studied clastic material contains fragments of crystalline rocks, and frequent clasts of Mesozoic to Palaeogene deep and shallow-water limestones. Numerous mica schists, scarce volcanites and granitoids as well as gneisses, quartzites and cataclasites were found in the group of crystalline exotic pebbles. Monazite ages of "exotic" mica-schist pebbles from the Tylicz, Zarzecze and Piwniczna-Mniszek sections document the Variscan 310±10 Ma age of metamorphic processes. The provenance of these exotic rocks could be connected with a remote source area located SE of the Magura Basin, which could be the NW part of the Dacia Mega Unit. The idea is strongly supported by palaeotransport directions from the SE, the absence of material derived from the Pieniny Klippen Belt, the presence of shallow water limestones, typical facies of the Median Dacides belt and metamorphic age distribution proved by monazite dating.

  7. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

    A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

  8. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  9. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  10. Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

    SciTech Connect

    David Petti

    2014-06-01

    Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO

  11. Supercell Depletion Studies for Prismatic High Temperature Reactors

    SciTech Connect

    J. Ortensi

    2012-10-01

    The traditional two-step method of analysis is not accurate enough to represent the neutronic effects present in the prismatic high temperature reactor concept. The long range coupling of the various regions in high temperature reactors poses a set of challenges that are not seen in either LWRs or fast reactors. Unlike LWRs, which exhibit large, localized effects, the dominant effects in PMRs are, for the most part, distributed over larger regions, but with lower magnitude. The 1-D in-line treatment currently used in pebble bed reactor analysis is not sufficient because of the 2-D nature of the prismatic blocks. Considerable challenges exist in the modeling of blocks in the vicinity of reflectors, which, for current small modular reactor designs with thin annular cores, include the majority of the blocks. Additional challenges involve the treatment of burnable poisons, operational and shutdown control rods. The use of a large domain for cross section preparation provides a better representation of the neutron spectrum, enables the proper modeling of BPs and CRs, allows the calculation of generalized equivalence theory parameters, and generates a relative power distribution that can be used in compact power reconstruction. The purpose of this paper is to quantify the effects of the reflector, burnable poison, and operational control rods on an LEU design and to delineate an analysis approach for the Idaho National Laboratory. This work concludes that the use of supercells should capture these long-range effects in the preparation of cross sections and along with a set of triangular meshes to treat BPs, and CRs a high fidelity neutronics computation is attainable.

  12. Brayton Cycle for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Oh, Chang H.; Moore, Richard L.

    2005-03-15

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others.The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  13. Brayton Cycle for High Temperature Gas-Cooled Reactors

    SciTech Connect

    Chang Oh

    2005-03-01

    This paper describes research on improving the Brayton cycle efficiency for a high-temperature gas-cooled reactor (HTGR). In this study, we are investigating the efficiency of an indirect helium Brayton cycle for the power conversion side of an HTGR power plant. A reference case based on a 250-MW(thermal) pebble bed HTGR was developed using helium gas as a working fluid in both the primary and power conversion sides. The commercial computer code HYSYS was used for process optimization. A numerical model using the Visual-Basic (V-B) computer language was also developed to assist in the evaluation of the Brayton cycle efficiency. Results from both the HYSYS simulation and the V-B model were compared with Japanese calculations based on the 300-MW(electric) Gas Turbine High-Temperature Reactor (GTHTR) that was developed by the Japan Atomic Energy Research Institute. After benchmarking our models, parametric investigations were performed to see the effect of important parameters on the cycle efficiency. We also investigated single-shaft versus multiple-shaft arrangements for the turbomachinery. The results from this study are applicable to other reactor concepts such as fast gas-cooled reactors, supercritical water reactors, and others. The ultimate goal of this study is to use other fluids such as supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency over that of the helium Brayton cycle. This study is in progress, and the results will be published in a subsequent paper.

  14. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  15. Hybrid adsorptive membrane reactor

    DOEpatents

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  16. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  17. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  18. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  19. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

    A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

  20. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  1. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  2. NEUTRONIC REACTOR MANIPULATING DEVICE

    DOEpatents

    Ohlinger, L.A.

    1962-08-01

    A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

  3. Multiphysics methods development for high temperature gas reactor analysis

    NASA Astrophysics Data System (ADS)

    Seker, Volkan

    Multiphysics computational methods were developed to perform design and safety analysis of the next generation Pebble Bed High Temperature Gas Cooled Reactors. A suite of code modules was developed to solve the coupled thermal-hydraulics and neutronics field equations. The thermal-hydraulics module is based on the three dimensional solution of the mass, momentum and energy equations in cylindrical coordinates within the framework of the porous media method. The neutronics module is a part of the PARCS (Purdue Advanced Reactor Core Simulator) code and provides a fine mesh finite difference solution of the neutron diffusion equation in three dimensional cylindrical coordinates. Coupling of the two modules was performed by mapping the solution variables from one module to the other. Mapping is performed automatically in the code system by the use of a common material mesh in both modules. The standalone validation of the thermal-hydraulics module was performed with several cases of the SANA experiment and the standalone thermal-hydraulics exercise of the PBMR-400 benchmark problem. The standalone neutronics module was validated by performing the relevant exercises of the PBMR-268 and PBMR-400 benchmark problems. Additionally, the validation of the coupled code system was performed by analyzing several steady state and transient cases of the OECD/NEA PBMR-400 benchmark problem.

  4. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  5. Comprehensive Thermal Hydraulics Research of the Very High Temperature Gas Cooled Reactor

    SciTech Connect

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; David Petti; Hyung Kang

    2010-10-01

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  6. Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

    SciTech Connect

    Massie, M.; Forsberg, C.; Forget, B.; Hu, L. W.

    2012-07-01

    A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

  7. Evaluation of the Accumulation of Trace Metals (as, U, CR, CU, PB, Zn) on Iron-Manganese Coatings on in Situ Stream Pebbles and Emplaced Substrates

    NASA Astrophysics Data System (ADS)

    Turpin, M. M.; Blake, J.; Crossey, L. J.; Ali, A.; Hansson, L.

    2015-12-01

    Exposure to trace metals (As, U, Cr, Cu, Pb, Zn) has potential negative health effects on human populations and wildlife. Geothermal waters often have elevated concentrations of trace elements and understanding the geochemical cycling of these elements can be challenging. Previous studies have utilized in situ stream pebbles and glass or ceramic substrates with iron-manganese oxide coatings to understand contamination and or chemical cycling. This project's main focus is to develop an ideal tracing method using adsorption onto substrate surfaces and to define key parameters that are necessary for the phenomenon of adsorption between trace metals and these surface coatings to occur. Sampling locations include the Jemez River and Rio San Antonio in the Jemez mountains, northern New Mexico. Both streams have significant geothermal inputs. Pebbles and cobbles were gathered from the active stream channel and 6mm glass beads and 2 X1 in. ceramic plates were placed in streams for three weeks to allow for coating accumulation. Factors such as leachate type, water pH, substrate type, coating accumulation period and leach time were all considered in this experiment. It was found that of the three leachates (aqua regia, 10% aqua regia and hydroxylamine), hydroxylamine was the most effective at leaching coatings without dissolving substrates. Samples leached with aqua regia and 10% aqua regia were found to lose weight and mass over the following 5, 7, and 10 day measurements. Glass beads were determined to be more effective than in stream pebbles as an accumulation substrate: coatings were more easily controlled and monitored. Samples leached with hydroxylamine for 5 hours and 72 hours showed little difference in their leachate concentrations, suggesting that leach time has little impact on the concentration of leachate samples. This research aims to find the best method for trace metal accumulation in streams to aid in understanding geochemical cycling.

  8. Tectonic setting of the pebble and other copper-gold-molybdenum porphyry deposits within the evolving middle cretaceous continental margin of Northwestern North America

    USGS Publications Warehouse

    Goldfarb, Richard J.; Anderson, Eric; Hart, Craig J.R.

    2013-01-01

    The Pebble Cu-Au-Mo deposit in southwestern Alaska, containing the largest gold resource of any known porphyry in the world, developed in a tectonic setting significantly different from that of the present-day. It is one of a series of metalliferous middle Cretaceous porphyritic granodiorite, quartz monzonite, and diorite bodies, evolved from lower crust and metasomatized lithospheric mantle melts, which formed along much of the length of the North American craton suture with the Peninsular-Alexander-Wrangellia arc. The porphyry deposits were emplaced within the northernmost two of a series of ca. 130 to 80 Ma flysch basins that define the suture, as well as into arc rocks immediately seaward of the two basins. Deposits include the ca. 100 to 90 Ma Pebble, Neacola, and other porphyry prospects along the Kahiltna basin-Peninsula terrane boundary, and the ca. 115 to 105 Ma Baultoff, Carl Creek, Horsfeld, Orange Hill, Bond Creek, and Chisna porphyries along the Nutzotin basin-Wrangellia terrane boundary.The porphyry deposits probably formed along the craton margin more than 1,000 km to the south of their present latitude. Palinspastic reconstructions of plate kinematics from this period are particularly difficult because magmatism overlaps the 119 to 83 Ma Cretaceous Normal Superchron, a period when sea-floor magnetic data are lacking. Our favored scenario is that ore formation broadly overlaps the cessation of sedimentation and contraction and the transition to a transpressional continental margin regime, such that the remnant ocean basins were converted to strike-slip basins. The basins and outboard Peninsular-Alexander-Wrangellia composite superterrane, which are all located seaward of the deep crustal Denali-Farewell fault system, were subjected to northerly dextral transpression for as long as perhaps 50 m.y., beginning at ca. 95 ± 10 Ma. The onset of this transpression was marked by development of the mineralized bodies along fault segments on the seaward side

  9. A comprehensive study of noble gases and nitrogen in "Hypatia", a diamond-rich pebble from SW Egypt

    NASA Astrophysics Data System (ADS)

    Avice, Guillaume; Meier, Matthias M. M.; Marty, Bernard; Wieler, Rainer; Kramers, Jan D.; Langenhorst, Falko; Cartigny, Pierre; Maden, Colin; Zimmermann, Laurent; Andreoli, Marco A. G.

    2015-12-01

    This is a follow-up study of a work by Kramers et al. (2013) on a very unusual diamond-rich rock fragment found in the area of south west Egypt in the south-western side of the Libyan Desert Glass strewn field. This pebble, called Hypatia, is composed of almost pure carbon. Transmission Electron Microscopy (TEM) and X-ray diffraction (XRD) results reveal that Hypatia is mainly made of defect-rich diamond containing lonsdaleite and multiple deformation bands. These characteristics are compatible with an impact origin on Earth and/or in space. We also analyzed concentrations and isotopic compositions of all five noble gases and nitrogen in several ∼mg sized Hypatia samples. These data confirm the conclusion by Kramers et al. (2013) that Hypatia is extra-terrestrial. The sample is relatively rich in trapped noble gases with an isotopic composition being close to the Q component found in many types of meteorites. 40Ar/36Ar ratios in individual steps are as low as 0.4 ± 0.3. Cosmic-ray produced "cosmogenic" 21Ne is present in concentrations corresponding to a nominal cosmic-ray exposure (CRE) age of roughly 0.1 Myr if produced in a typical meter-sized meteoroid. Such an atypically low nominal CRE age suggests high shielding in a considerably larger body. In addition to the Xe-Q composition, an excess of radiogenic 129Xe (from the decay of short-lived radioactive 129I) is observed (129Xe /132Xe = 1.18 + / - 0.03). Two isotopically distinct N components are present, an isotopically heavy component (δ15N ∼ + 20 ‰) released at low temperatures and a major isotopically light component (δ15N ∼ - 110 ‰) at higher temperatures. This disequilibrium in N suggests that the diamonds in Hypatia were formed in space rather than upon impact on Earth (δN15atm = 0 ‰). All our data are broadly consistent with concentrations and isotopic compositions of noble gases in at least three different types of carbon-rich meteoritic materials: carbon-rich veins in ureilites

  10. Overview of Indian activities on fusion reactor materials

    NASA Astrophysics Data System (ADS)

    Banerjee, Srikumar

    2014-12-01

    This paper on overview of Indian activities on fusion reactor materials describes in brief the efforts India has made to develop materials for the first wall of a tokamak, its blanket and superconducting magnet coils. Through a systematic and scientific approach, India has developed and commercially produced reduced activation ferritic/martensitic (RAFM) steel that is comparable to Eurofer 97. Powder of low activation ferritic/martensitic oxide dispersion strengthened steel with characteristics desired for its application in the first wall of a tokamak has been produced on the laboratory scale. V-4Cr-4Ti alloy was also prepared in the laboratory, and kinetics of hydrogen absorption in this was investigated. Cu-1 wt%Cr-0.1 wt%Zr - an alloy meant for use as heat transfer elements for hypervapotrons and heat sink for the first wall - was developed and characterized in detail for its aging behavior. The role of addition of a small quantity of Zr in its improved fatigue performance was delineated, and its diffusion bonding with both W and stainless steel was achieved using Ni as an interlayer. The alloy was produced in large quantities and used for manufacturing both the heat transfer elements and components for the International Thermonuclear Experimental Reactor (ITER). India has proposed to install and test a lead-lithium cooled ceramic breeder test blanket module (LLCB-TBM) at ITER. To meet this objective, efforts have been made to produce and characterize Li2TiO3 pebbles, and also improve the thermal conductivity of packed beds of these pebbles. Liquid metal loops have been set up and corrosion behavior of RAFM steel in flowing Pb-Li eutectic has been studied in the presence as well as absence of magnetic fields. To prevent permeation of tritium and reduce the magneto-hydro-dynamic drag, processes have been developed for coating alumina on RAFM steel. Apart from these activities, different approaches being attempted to make the U-shaped first wall of the TBM box

  11. Analysis of two-dimensional shapes by principal component score descriptors: geological interpretations from sand grains, pebbles, benthic foraminifera, and bivalve mollusks

    SciTech Connect

    Parks, J.M.

    1987-05-01

    Computerized quantitative shape analysis provides useful geological information not readily obtained in other ways. Principal Components Analysis (PCA) of properly rotated images reduces digitized outlines to a few shape descriptors. R-mode PC loadings, displayed graphically, exhibit the distinctive components of shape (elongation, triangularity, rectangularity, etc) in different orientations. Q-mode estimated PC scores are the shape descriptors for individual objects. Six shape descriptors are adequate to characterize typical geological outline shapes, such as silhouettes of sand grains, pebbles, and fossils. The original outlines are reconstructed using these shape descriptors as proportions for recombining the PC loadings. Proportions and rates of sand mixing from two sources are revealed by shape analysis of populations of sand grains from the Kansas and Missouri rivers sampled above and below their confluence. Unmixing (differential sorting during transport) is revealed by gradual shape changes in sands sampled along 330 mi of the Rio Grande (Del Rio, Texas, to the Gulf of Mexico). Pebbles from the Jackson Hole, Wyoming, area are readily identified as to fluvial or glacial origin by quantified shape. Outline shapes of benthic foraminifera from Maryland Miocene assemblages are classified by cluster and discriminant analyses of PC scores into 20 or more morphological types. Relative proportions of each morpho-type in stratigraphic samples are statistically correlated with independent paleoenvironmental indicators. Intra- and inter-specific changes in shapes of several genera of middle Miocene bivalves from Maryland show three distinct patterns through time: minor irregular changes (= stasis.); abrupt jumps (= punctuated equilibria.); and gradual trends (= gradualism.).

  12. Storm recovery on two Italian coarse-grained beaches: a comparison between a mixed sand and gravel and a pebble beach

    NASA Astrophysics Data System (ADS)

    Bertoni, Duccio; Grottoli, Edoardo; Ciavola, Paolo; Sarti, Giovanni; Pozzebon, Alessandro

    2014-05-01

    High energy events emphasize beach erosion processes, sometimes leading to huge volume deficits not balanced by recovery under fair-weather conditions. In this scenario, artificial replenishments are frequently used as a form of coastal protection with large volumes of sediments re-injected in the system without strongly altering the environment as it happens with hard structures. Since climate change is expected to accentuate in the near future erosion effects, the need to artificially feed beaches is likely to increase. Gravel and pebbles are more and more often used as beach fill, on some occasions replacing sandy sediments. That was the case for two beaches located at either sides of the Italian Peninsula (Portonovo, Adriatic Sea; Marina di Pisa, Ligurian Sea), which constitute the study area of the present research. Portonovo is a 500 m-long mixed sand and gravel beach with a significant pebble-sized content (about 40%), unloaded on the beach during multiple replenishments. Marina di Pisa is an artificial, 180 m-long beach, mainly composed of 40-to-90 mm pebbles; it was built in 2008 as a part of a larger protection scheme. Groins or headlands that prevent any sediment exchange with adjacent areas bound both beaches. Periodic topographic surveys were carried out to evaluate the response of these human-altered beaches to high-energy events. The topographic surveys, undertaken with a DGPS-RTK instrument along cross-shore transects (from the landward end of the backshore to about 1.5 m depth seaward), were done following intense storm events occurred during the time period of the research. Transects were done out every 10 m along the entire length of the beaches. Prior to the first topographic survey, a sediment tracing experiment was set up as a form of control of the results provided by the geomorphologic analysis. Pebbles directly sampled from the beaches were marked by means of the RFID technology and injected back all along the beachface. As expected

  13. Hybrid plasmachemical reactor

    SciTech Connect

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  14. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  15. Attrition reactor system

    SciTech Connect

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  16. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  17. Reactor System Transient Code.

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  18. A pebble curiosity

    NASA Astrophysics Data System (ADS)

    Gardner, Martin

    2000-01-01

    For this experiment you need a supply of marbles of two different sizes. Any kind of small spheres will do also, such as spherical bubble gum and candy. Assuming you are using marbles, fill a glass to the brim with the small marbles, and fill another glass to the brim with the larger marbles. Which glass has the larger volume of empty space?

  19. More Pebble Puzzles.

    ERIC Educational Resources Information Center

    Gibbs, William M.

    This booklet is a collection of puzzles, investigations, and games. They are designed to be used with large objects such as tins or stones and diagrams marked on the ground. The children are to be encouraged to use an experimental, trial-and-error approach at first, and then develop methods of solution. (MNS)

  20. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  1. High solids fermentation reactor

    DOEpatents

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  2. Efficient Silicon Reactor

    NASA Technical Reports Server (NTRS)

    Bates, H. E.; Hill, D. M.; Jewett, D. N.

    1983-01-01

    High-purity silicon efficiently produced and transferred by continuous two-cycle reactor. New reactor operates in relatively-narrow temperature rate and uses large surfaces area to minimize heat expenditure and processing time in producing silicon by hydrogen reduction of trichlorosilane. Two cycles of reactor consists of silicon production and removal.

  3. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  4. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  5. NEUTRONIC REACTOR SHIELDING

    DOEpatents

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  6. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  7. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  8. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    SciTech Connect

    Not Available

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

  9. Nuclear reactor overflow line

    DOEpatents

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  10. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  11. Reactor water cleanup system

    DOEpatents

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  12. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  13. High temperature reactors

    NASA Astrophysics Data System (ADS)

    Dulera, I. V.; Sinha, R. K.

    2008-12-01

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements.

  14. Spinning fluids reactor

    DOEpatents

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  15. Nuclear Reactor Physics

    NASA Astrophysics Data System (ADS)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  16. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  17. Modelling of biofilm reactors

    SciTech Connect

    Rodrigues, A.; Grasmick, A.; Elmaleh, S.

    1982-10-01

    Comprehensive models of biofilm reactors are developed. Model I assumes a zero-order reaction of a limiting substrate and a diffusional mass transport through the biofilm; in the diffusion-controlled regime the model is fully characterized by one parameter alpha. From this model the conversion of substrate or reactor efficiency can be calculated, for continuously stirred tank reactors (CSTRs) and plug flow reactors respectively, as follows: EA = )alpha(alpha + 2)) 1/2 - alpha; and Ep = (2 alpha) 1/2 - alpha/2: Validation of the model is tested for different experimental systems. Model II includes liquid film mass transfer resistance. The conversion gap between plug flow reactors and CSTRs is always lower than 25% and, as a first approximation, the biofilm reactor design does not then require accurate residence time distribution measurements. (Refs. 23).

  18. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  19. Thermionic space reactors overview

    SciTech Connect

    Wetch, J.R.; Britt, E.J.; Fitzpatrick, G.O.; Rasor, N.S.

    1983-08-01

    The multi-national development of thermionic reactor systems is summarized in the context of the past general space nuclear reactor program and the recent renewed interest in space nuclear power. Comparison of various alternate reactor space power systems indicates that only the in-core thermionic reactor approach has the performance and growth potential required to provide the power levels potentially needed for shuttle-launchable systems by the year 2000 at reactor coolant and system temperatures that are near the current state-of-the-art. It is concluded that all shuttle- launchable high power space reactor systems require high-temperature, long-endurance nuclear fuels, and that high priority characterization and development of such fuels is essential to successfully realize power systems that can enable the space missions presently being considered.

  20. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Dreffin, R.S.

    1959-12-15

    A control means for a nuclear reactor is described. Particularly a device extending into the active portion of the reactor consisting of two hollow elements coaxially disposed and forming a channel therebetween, the cross sectional area of the channel increasing from each extremity of the device towards the center thereof. An element of neutron absorbing material is slidably positionable within the inner hollow element and a fluid reactor poison is introduced into the channel defined by the two hollow elements.

  1. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  2. Improved vortex reactor system

    DOEpatents

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  3. THERMAL NEUTRONIC REACTOR

    DOEpatents

    Spinrad, B.I.

    1960-01-12

    A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

  4. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  5. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  6. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

  7. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  8. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    SciTech Connect

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-06-19

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  9. Implications of Graphite Radiation Damage on the Neutronic, Operational, and Safety Aspects of Very High Temperature Reactors

    SciTech Connect

    Hawari, Ayman I

    2011-08-30

    In both the prismatic and pebble bed designs of Very High Temperature Reactors (VHTR), the graphite moderator is expected to reach exposure levels of 1021 to 1022 n/cm2 over the lifetime of the reactor. This exposure results in damage to the graphite structure. In this work, molecular dynamic and ab initio molecular static calculations will be used to: 1) simulate radiation damage in graphite under various irradiation and temperature conditions, 2) generate the thermal neutron scattering cross sections for damaged graphite, and 3) examine the resulting microstructure to identify damage formations that may produce the high-temperature Wigner effect. The impact of damage on the neutronic, operational and safety behavior of the reactor will be assessed using reactor physics calculations. In addition, tests will be performed on irradiated graphite samples to search for the high-temperature Wigner effect, and phonon density of states measurements will be conducted to quantify the effect on thermal neutron scattering cross sections using these samples.

  10. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  11. Polymerization Reactor Engineering.

    ERIC Educational Resources Information Center

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  12. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. ); Lineberry, M.J. )

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

  13. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  14. Operating US power reactors

    SciTech Connect

    Silver, E.G.

    1982-07-01

    The operation of US power reactors during March and April 1982 is summarized. Events of special note are discussed in the text, and the operational performance of all licensed power reactors is presented. These data are taken from the monthly Operating Units Status Report prepared by the Nuclear Regulatory Commission (NRC).

  15. Compact torsatron reactors

    SciTech Connect

    Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

    1988-05-01

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R/sub 0/ = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R/sub 0/ approx. = 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs.

  16. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  17. Status of French reactors

    SciTech Connect

    Ballagny, A.

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  18. Reactor neutrino monitoring

    NASA Astrophysics Data System (ADS)

    Lhuillier, D.

    2009-03-01

    Nuclear reactors are the most intense man-controlled sources of antineutrinos and as such have hosted number of key physics experiments, from the antineutrino discovery to modern oscillation measurements. At the present time, both detection technology and understanding of fundamental physics are mature enough to think about antineutrinos as a new tool for reactor monitoring. We describe below how antineutrinos can provide online information on reactor operation and amount of plutonium accumulated in the core. Reactors are the only sources of plutonium on earth and this element can be chemically separated from the rest of the nuclear fuel and diverted into nuclear weapons. We present in the next sections the unique features antineutrino detectors could provide to safeguards agencies such as IAEA. We review the worldwide efforts to develop small ( 1m scale) antineutrino detectors dedicated to automated and non-intrusive reactor monitoring.

  19. Nuclear reactor control column

    DOEpatents

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  20. Slurry reactor design studies

    SciTech Connect

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  1. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  2. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Galvez, Cristhian

    2011-12-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data. This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study. A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the

  3. Nuclear reactor reflector

    DOEpatents

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  4. Nuclear reactor reflector

    DOEpatents

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  5. Microfluidic electrochemical reactors

    DOEpatents

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  6. CONTROL FOR NEUTRONIC REACTOR

    DOEpatents

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  7. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  8. COOLED NEUTRONIC REACTOR

    DOEpatents

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  9. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  10. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  11. Fast Breeder Reactor studies

    SciTech Connect

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  12. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  13. Packed Bed Reactor Experiment

    NASA Video Gallery

    The purpose of the Packed Bed Reactor Experiment in low gravity is to determine how a mixture of gas and liquid flows through a packed bed in reduced gravity. A packed bed consists of a metal pipe ...

  14. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Daniels, F.

    1961-10-24

    A reactor core, comprised of vertical stacks of hexagonal blocks of beryllium oxide having axial cylindrical apertures extending therethrough and cylindrical rods of a sintered mixture of uranium dioxide and beryllium oxide, is described. (AEC)

  15. NEUTRONIC REACTOR FUEL COMPOSITION

    DOEpatents

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  16. Results of a Neutronic Simulation of HTR-Proteus Core 4.2 using PEBBED and other INL Reactor Physics Tools: FY-09 Report

    SciTech Connect

    Hans D. Gougar

    2009-08-01

    The Idaho National Laboratory’s deterministic neutronics analysis codes and methods were applied to the computation of the core multiplication factor of the HTR-Proteus pebble bed reactor critical facility. A combination of unit cell calculations (COMBINE-PEBDAN), 1-D discrete ordinates transport (SCAMP), and nodal diffusion calculations (PEBBED) were employed to yield keff and flux profiles. Preliminary results indicate that these tools, as currently configured and used, do not yield satisfactory estimates of keff. If control rods are not modeled, these methods can deliver much better agreement with experimental core eigenvalues which suggests that development efforts should focus on modeling control rod and other absorber regions. Under some assumptions and in 1D subcore analyses, diffusion theory agrees well with transport. This suggests that developments in specific areas can produce a viable core simulation approach. Some corrections have been identified and can be further developed, specifically: treatment of the upper void region, treatment of inter-pebble streaming, and explicit (multiscale) transport modeling of TRISO fuel particles as a first step in cross section generation. Until corrections are made that yield better agreement with experiment, conclusions from core design and burnup analyses should be regarded as qualitative and not benchmark quality.

  17. Molten metal reactors

    DOEpatents

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  18. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  19. Future reactor experiments

    SciTech Connect

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  20. Assessments of Water Ingress Accidents in a Modular High-Temperature Gas-Cooled Reactor

    SciTech Connect

    Zhang Zuoyi; Dong Yujie; Scherer, Winfried

    2005-03-15

    Severe water ingress accidents in the 200-MW HTR-module were assessed to determine the safety margins of modular pebble-bed high-temperature gas-cooled reactors (HTR-module). The 200-MW HTR-module was designed by Siemens under the criteria that no active safety protection systems were necessary because of its inherent safe nature. For simulating the behavior of the HTR-module during severe water ingress accidents, a water, steam, and helium multiphase cavity model was developed and implemented in the dynamic simulator for nuclear power plants (DSNP) simulation system. Comparisons of the DSNP simulations incorporating these models with experiments and with calculations using the time-dependent neutronics and temperature dynamics code were made to validate the simulation. The analysis of the primary circuit showed that the maximum water concentration increase in the reactor core was <0.3 kg/(m{sup 3}s). The water vaporization in the steam generator and characteristics of water transport from the steam generator to the reactor core would reduce the rate of water ingress into the reactor core. The analysis of a full cavitation of the feedwater pump showed that if the secondary circuit could be depressurized, the feedwater pump would be stopped by the full cavitation. This limits the water transported from the deaerator to the steam generator. A comprehensive simulation of the HTR-module power plant showed that the water inventory in the primary circuit was limited to {approx}3000 kg. The nuclear reactivity increase caused by the water ingress would lead to a fast power excursion, which would be inherently counterbalanced by negative feedback effects. The integrity of the fuel elements, because the safety-relevant temperature limit of 1600 deg. C is not reached in any case, is not challenged.

  1. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  2. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  3. F Reactor Inspection

    SciTech Connect

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  4. F Reactor Inspection

    ScienceCinema

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2016-07-12

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  5. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  6. NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

    DOEpatents

    West, J.M.; Weills, J.T.

    1960-03-15

    A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

  7. Morphological and sedimentary impacts and recovery on a mixed sandy to pebbly seabed exposed to marine aggregate extraction (Eastern English Channel, France)

    NASA Astrophysics Data System (ADS)

    Le Bot, Sophie; Lafite, Robert; Fournier, Matthieu; Baltzer, Agnès; Desprez, Michel

    2010-10-01

    The impacts of marine aggregate dredging and the recovery levels, 1 and 10 years after cessation of dredging, are investigated for a commercial dredging site. The site consists of mixed sandy to pebbly sediments. Dredging intensity is low (generally less than 1 h/ha/year) and heterogeneous in space and time, allowing to distinguish adjacent areas of active extraction, fallow, recovery and overflow deposition. Surveys were conducted using a combination of acoustic (side-scan sonar), video and grab sampling techniques. Multivariate analysis has been utilized to define a seabed morphological and sedimentary zonation that provides a good habitat description and mapping. Sedimentary patchiness characterized by grain-size increased variability is identified as a feature indicative of seabed disturbance due to dredging activities. In the vicinity of the dredging areas, fine-sand content greater than 10% is indicative of overflow deposition resulting from dredging activities while coarse sand is found to be covering areas previously dredged but now recovered. Furrows created by dredging operations collapse within one year, and generally fully disappear within 10 years. They are usually covered by sands and bedforms, indicative of active sediment transport, are formed. This study confirms that a low dredging intensity strategy is recommended for mixed sediment areas under moderate hydrodynamic energy regimes as this of the study area. This allows natural recovery of the seabed to take place within a period of 10 years.

  8. A novel scaffold protein, TANC, possibly a rat homolog of Drosophila rolling pebbles (rols), forms a multiprotein complex with various postsynaptic density proteins.

    PubMed

    Suzuki, Tatsuo; Li, Weidong; Zhang, Jing-Ping; Tian, Qing-Bao; Sakagami, Hiroyuki; Usuda, Nobuteru; Usada, Nobuteru; Kondo, Hisatake; Fujii, Toshihiro; Endo, Shogo

    2005-01-01

    We cloned from the rat brain a novel gene, tanc (GenBank Accession No. AB098072), which encoded a protein containing three tetratricopeptide repeats (TPRs), ten ankyrin repeats and a coiled-coil region, and is possibly a rat homolog of Drosophila rolling pebbles (rols). The tanc gene was expressed widely in the adult rat brain. Subcellular distribution, immunohistochemical study of the brain and immunocytochemical studies of cultured neuronal cells indicated the postsynaptic localization of TANC protein of 200 kDa. Pull-down experiments showed that TANC protein bound PSD-95, SAP97, and Homer via its C-terminal PDZ-binding motif, -ESNV, and fodrin via both its ankyrin repeats and the TPRs together with the coiled-coil domain. TANC also bound the alpha subunit of Ca2+/calmodulin-dependent protein kinase II. An immunoprecipitation study showed TANC association with various postsynaptic proteins, including guanylate kinase-associated protein (GKAP), alpha-internexin, and N-methyl-D-aspartate (NMDA)-type glutamate receptor 2B and AMPA-type glutamate receptor (GluR1) subunits. These results suggest that TANC protein may work as a postsynaptic scaffold component by forming a multiprotein complex with various postsynaptic density proteins. PMID:15673434

  9. An Advanced Integrated Diffusion/Transport Method for the Design, Analysis and Optimization of the Very-High-Temperature Reactors

    SciTech Connect

    Farzad Rahnema; Dingkang Zhang; Abderrafi Ougouag; Frederick Gleicher

    2011-04-04

    The main objective of this research is to develop an integrated diffusion/transport (IDT) method to substantially improve the accuracy of nodal diffusion methods for the design and analysis of Very High Temperature Reactors (VHTR). Because of the presence of control rods in the reflector regions in the Pebble Bed Reactor (PBR-VHTR), traditional nodal diffusion methods do not accurately model these regions, within which diffusion theory breaks down in the vicinity of high neutron absorption and steep flux gradients. The IDT method uses a local transport solver based on a new incident flux response expansion method in the controlled nodes. Diffusion theory is used in the rest of the core. This approach improves the accuracy of the core solution by generating transport solutions of controlled nodes while maintaining computational efficiency by using diffusion solutions in nodes where such a treatment is sufficient. The transport method is initially developed and coupled to the reformulated 3-D nodal diffusion model in the CYNOD code for PBR core design and fuel cycle analysis. This method is also extended to the prismatic VHTR. The new method accurately captures transport effects in highly heterogeneous regions with steep flux gradients. The calculations of these nodes with transport theory avoid errors associated with spatial homogenization commonly used in diffusion methods in reactor core simulators

  10. Deep Burn: Development of Transuranic Fuel for High-Temperature Helium-Cooled Reactors- Monthly Highlights September 2010

    SciTech Connect

    Snead, Lance Lewis; Besmann, Theodore M; Collins, Emory D; Bell, Gary L

    2010-10-01

    The DB Program monthly highlights report for August 2010, ORNL/TM-2010/184, was distributed to program participants by email on September 17. This report discusses: (1) Core and Fuel Analysis - (a) Core Design Optimization in the HTR (high temperature helium-cooled reactor) Prismatic Design (Logos), (b) Core Design Optimization in the HTR Pebble Bed Design (INL), (c) Microfuel analysis for the DB HTR (INL, GA, Logos); (2) Spent Fuel Management - (a) TRISO (tri-structural isotropic) repository behavior (UNLV), (b) Repository performance of TRISO fuel (UCB); (3) Fuel Cycle Integration of the HTR (high temperature helium-cooled reactor) - Synergy with other reactor fuel cycles (GA, Logos); (4) TRU (transuranic elements) HTR Fuel Qualification - (a) Thermochemical Modeling, (b) Actinide and Fission Product Transport, (c) Radiation Damage and Properties; (5) HTR Spent Fuel Recycle - (a) TRU Kernel Development (ORNL), (b) Coating Development (ORNL), (c) Characterization Development and Support, (d) ZrC Properties and Handbook; and (6) HTR Fuel Recycle - (a) Graphite Recycle (ORNL), (b) Aqueous Reprocessing, (c) Pyrochemical Reprocessing METROX (metal recovery from oxide fuel) Process Development (ANL).

  11. REACTOR GROUT THERMAL PROPERTIES

    SciTech Connect

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  12. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  13. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    SciTech Connect

    Geelhood, Kenneth J.; Rishel, Jeremy P.

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  14. EBT reactor analysis

    SciTech Connect

    Uckan, N. A.; Jaeger, E. F.; Santoro, R. T.; Spong, D. A.; Uckan, T.; Owen, L. W.; Barnes, J. M.; McBride, J. B.

    1983-08-01

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with <..beta../sub core/> approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m/sup 2/, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density.

  15. REACTOR AND NOVEL METHOD

    DOEpatents

    Young, G.J.; Ohlinger, L.A.

    1958-06-24

    A nuclear reactor of the type which uses a liquid fuel and a method of controlling such a reactor are described. The reactor is comprised essentially of a tank for containing the liquid fuel such as a slurry of discrete particles of fissionnble material suspended in a heavy water moderator, and a control means in the form of a disc of neutron absorbirg material disposed below the top surface of the slurry and parallel thereto. The diameter of the disc is slightly smaller than the diameter of the tank and the disc is perforated to permit a flow of the slurry therethrough. The function of the disc is to divide the body of slurry into two separate portions, the lower portion being of a critical size to sustain a nuclear chain reaction and the upper portion between the top surface of the slurry and the top surface of the disc being of a non-critical size. The method of operation is to raise the disc in the reactor until the lower portion of the slurry has reached a critical size when it is desired to initiate the reaction, and to lower the disc in the reactor to reduce the size of the lower active portion the slurry to below criticality when it is desired to stop the reaction.

  16. Methanation assembly using multiple reactors

    DOEpatents

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  17. Mirror Advanced Reactor Study (MARS): executive summary and overview

    SciTech Connect

    Logan, B.G.; Perkins, L.J.; Gordon, J.D.

    1984-07-01

    Two self-consistent MARS configurations are discussed - a 1200-MWe commercial electricity-generating plant and a synguels-generating plant that produces hydrogen with an energy equivalent to 26,000 barrels of oil per day. The MARS machine emphasizes the attractive features of the tandem mirror concept, including steady-state operation, a small-diameter high-beta plasma, a linear central cell with simple low-maintenance blankets, low first-wall heat fluxes (<10 W/cm/sup 2/), no driven plasma currents or associated disruptions, natural halo impurity diversion, and direct conversion of end-loss charged-particle power. The MARS electric plant produces 2600 MW of fusion power in a 130-m-long central cell. Advanced tandem-mirror plasma-engineering concepts, a high-efficiency liquid lithium-lead (Li/sub 17/Pb/sub 83/) blanket, and efficient direct electrical conversion of end loss power combine to produce a high net plant efficiency of 36%. With a total capital cost of $2.9 billion (constant 1983 dollars), the MARS electric plant produces busbar electricity at approx. 7 cents/kW-hour. The MARS synfuels plant produces 3500 MW of fusion power in a 150-m-long central cell. A helium-gas-cooled silicon carbide pebble-bed blanket provides high-temperature (1000/sup 0/C) heat to a thermochemical water-splitting cycle and the resulting hydrogen is catalytically converted to methanol for distribution. With a total capital cost of $3.6 billion (constant 1983 dollars), the synfuels plant produces methanol fuel at about $1.7/gal. The major features of the MARS reactor include sloshing-ion thermal barrier plugs for efficient plasma confinement, a high efficiency blanket, high-field (24-T) choke cells, drift pumping for trapped plasma species, quasi-optical electron-cyclotron resonant heating (ECRH) systems, and a component gridless direct converter.

  18. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    NASA Astrophysics Data System (ADS)

    Zweibaum, Nicolas

    The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). (Abstract shortened by UMI.).

  19. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  20. Reactor for exothermic reactions

    DOEpatents

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  1. A NEUTRONIC REACTOR

    DOEpatents

    Luebke, E.A.; Vandenberg, L.B.

    1959-09-01

    A nuclear reactor for producing thermoelectric power is described. The reactor core comprises a series of thermoelectric assemblies, each assembly including fissionable fuel as an active element to form a hot junction and a thermocouple. The assemblies are disposed parallel to each other to form spaces and means are included for Introducing an electrically conductive coolant between the assemblies to form cold junctions of the thermocouples. An electromotive force is developed across the entire series of the thermoelectric assemblies due to fission heat generated in the fuel causing a current to flow perpendicular to the flow of coolant and is distributed to a load outside of the reactor by means of bus bars electrically connected to the outermost thermoelectric assembly.

  2. Dynamic bed reactor

    SciTech Connect

    Stormo, K.E.

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix. 27 figs.

  3. Reactor for exothermic reactions

    DOEpatents

    Smith, Jr., Lawrence A.; Hearn, Dennis; Jones, Jr., Edward M.

    1993-01-01

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  4. Dynamic bed reactor

    DOEpatents

    Stormo, Keith E.

    1996-07-02

    A dynamic bed reactor is disclosed in which a compressible open cell foam matrix is periodically compressed and expanded to move a liquid or fluid through the matrix. In preferred embodiments, the matrix contains an active material such as an enzyme, biological cell, chelating agent, oligonucleotide, adsorbent or other material that acts upon the liquid or fluid passing through the matrix. The active material may be physically immobilized in the matrix, or attached by covalent or ionic bonds. Microbeads, substantially all of which have diameters less than 50 microns, can be used to immobilize the active material in the matrix and further improve reactor efficiency. A particularly preferred matrix is made of open cell polyurethane foam, which adsorbs pollutants such as polychlorophenol or o-nitrophenol. The reactors of the present invention allow unidirectional non-laminar flow through the matrix, and promote intimate exposure of liquid reactants to active agents such as microorganisms immobilized in the matrix.

  5. Heat dissipating nuclear reactor

    DOEpatents

    Hunsbedt, Anstein; Lazarus, Jonathan D.

    1987-01-01

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  6. Heat dissipating nuclear reactor

    DOEpatents

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  7. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-08-01

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  8. Compact reactor design automation

    NASA Technical Reports Server (NTRS)

    Nassersharif, Bahram; Gaeta, Michael J.

    1991-01-01

    A conceptual compact reactor design automation experiment was performed using the real-time expert system G2. The purpose of this experiment was to investigate the utility of an expert system in design; in particular, reactor design. The experiment consisted of the automation and integration of two design phases: reactor neutronic design and fuel pin design. The utility of this approach is shown using simple examples of formulating rules to ensure design parameter consistency between the two design phases. The ability of G2 to communicate with external programs even across networks provides the system with the capability of supplementing the knowledge processing features with conventional canned programs with possible applications for realistic iterative design tools.

  9. REACTOR CONTROL DEVICE

    DOEpatents

    Graham, R.H.

    1962-09-01

    A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

  10. Merchant Marine Ship Reactor

    DOEpatents

    Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

    1961-05-01

    A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

  11. MERCHANT MARINE SHIP REACTOR

    DOEpatents

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  12. Advanced light water reactor requirements document: Chapter 4, Reactor systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the following for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR): reactor pressure vessel, nozzles and safe-ends, reactor internals, in-vessel portions of fluid systems (including reactor internal pumps (Emergency Core Cooling System (ECCS) piping and spargers), nuclear fuel, and the control rods and control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection and testing are also covered.

  13. Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

  14. Plug Flow Reactor Simulator

    SciTech Connect

    Larson, Richard S.

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position, and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.

  15. NEUTRONIC REACTOR CONTROL ELEMENT

    DOEpatents

    Newson, H.W.

    1960-09-13

    A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

  16. MEANS FOR SHIELDING REACTORS

    DOEpatents

    Garrison, W.M.; McClinton, L.T.; Burton, M.

    1959-03-10

    A reactor of the heterageneous, heavy water moderated type is described. The reactor is comprised of a plurality of vertically disposed fuel element tubes extending through a tank of heavy water moderator and adapted to accommodate a flow of coolant water in contact with the fuel elements. A tank containing outgoing coolant water is disposed above the core to function is a radiation shield. Unsaturated liquid hydrocarbon is floated on top of the water in the shield tank to reduce to a minimum the possibility of the occurrence of explosive gaseous mixtures resulting from the neutron bombardment of the water in the shield tank.

  17. Perspectives on reactor safety

    SciTech Connect

    Haskin, F.E.; Camp, A.L.

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  18. Fusion reactor pumped laser

    DOEpatents

    Jassby, Daniel L.

    1988-01-01

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  19. ARIES tokamak reactor study

    SciTech Connect

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein.

  20. Plug Flow Reactor Simulator

    1996-07-30

    PLUG is a computer program that solves the coupled steady state continuity, momentum, energy, and species balance equations for a plug flow reactor. Both homogeneous (gas-phase) and heterogenous (surface) reactions can be accommodated. The reactor may be either isothermal or adiabatic or may have a specified axial temperature or heat flux profile; alternatively, an ambient temperature and an overall heat-transfer coefficient can be specified. The crosssectional area and surface area may vary with axial position,more » and viscous drag is included. Ideal gas behavior and surface site conservation are assumed.« less

  1. Particle bed reactor modeling

    NASA Technical Reports Server (NTRS)

    Sapyta, Joe; Reid, Hank; Walton, Lew

    1993-01-01

    The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

  2. Nuclear reactor apparatus

    DOEpatents

    Wade, Elman E.

    1978-01-01

    A lifting, rotating and sealing apparatus for nuclear reactors utilizing rotating plugs above the nuclear reactor core. This apparatus permits rotation of the plugs to provide under the plug refueling of a nuclear core. It also provides a means by which positive top core holddown can be utilized. Both of these operations are accomplished by means of the apparatus lifting the top core holddown structure off the nuclear core while stationary, and maintaining this structure in its elevated position during plug rotation. During both of these operations, the interface between the rotating member and its supporting member is sealingly maintained.

  3. Fast quench reactor method

    SciTech Connect

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.; Berry, Ray A.

    1999-01-01

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  4. Fast quench reactor method

    DOEpatents

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  5. THERMAL NUCLEAR REACTOR

    DOEpatents

    Fenning, F.W.; Jackson, R.F.

    1957-09-24

    Nuclear reactors of the graphite moderated air cooled type in which canned slugs or rods of fissile material are employed are discussed. Such a reactor may be provided with a means for detecting dust particles in the exhausted air. The means employed are lengths of dust absorbent cord suspended in vertical holes in the shielding structure above each vertical coolant flow channel to hang in the path of the cooling air issuing from the channels, and associated spindles and drive motors for hauling the cords past detectors, such as Geiger counters, for inspecting the cords periodically. This design also enables detecting the individual channel in which a fault condition may have occurred.

  6. Diagnostics for hybrid reactors

    SciTech Connect

    Orsitto, Francesco Paolo

    2012-06-19

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  7. Myosin heavy chain-like localizes at cell contact sites during Drosophila myoblast fusion and interacts in vitro with Rolling pebbles 7

    SciTech Connect

    Bonn, Bettina R.; Rudolf, Anja; Hornbruch-Freitag, Christina; Daum, Gabor; Kuckwa, Jessica; Kastl, Lena; Buttgereit, Detlev; Renkawitz-Pohl, Renate

    2013-02-15

    Besides representing the sarcomeric thick filaments, myosins are involved in many cellular transport and motility processes. Myosin heavy chains are grouped into 18 classes. Here we show that in Drosophila, the unconventional group XVIII myosin heavy chain-like (Mhcl) is transcribed in the mesoderm of embryos, most prominently in founder cells (FCs). An ectopically expressed GFP-tagged Mhcl localizes in the growing muscle at cell–cell contacts towards the attached fusion competent myoblast (FCM). We further show that Mhcl interacts in vitro with the essential fusion protein Rolling pebbles 7 (Rols7), which is part of a protein complex established at cell contact sites (Fusion-restricted Myogenic-Adhesive Structure or FuRMAS). Here, branched F-actin is likely needed to widen the fusion pore and to integrate the myoblast into the growing muscle. We show that the localization of Mhcl is dependent on the presence of Rols7, and we postulate that Mhcl acts at the FuRMAS as an actin motor protein. We further show that Mhcl deficient embryos develop a wild-type musculature. We thus propose that Mhcl functions redundantly to other myosin heavy chains in myoblasts. Lastly, we found that the protein is detectable adjacent to the sarcomeric Z-discs, suggesting an additional function in mature muscles. - Highlights: ► The class XVIII myosin encoding gene Mhcl is transcribed in the mesoderm. ► Mhcl localization at contact sites of fusing myoblasts depends on Rols7. ► Mhcl interacts in vitro with Rols7 which is essential for myogenesis. ► Functional redundancy with other myosins is likely as mutants show no muscle defects. ► Mhcl localizes adjacent to Z-discs of sarcomeres and might support muscle integrity.

  8. Integrated geophysical imaging of a concealed mineral deposit: a case study of the world-class Pebble porphyry deposit in southwestern Alaska

    USGS Publications Warehouse

    Shah, Anjana K.; Bedrosian, Paul A.; Anderson, Eric D.; Kelley, Karen D.; Lang, James

    2013-01-01

    We combined aeromagnetic, induced polarization, magnetotelluric, and gravity surveys as well as drillhole geologic, alteration, magnetic susceptibility, and density data for exploration and characterization of the Cu-Au-Mo Pebble porphyry deposit. This undeveloped deposit is almost completely concealed by postmineralization sedimentary and volcanic rocks, presenting an exploration challenge. Individual geophysical methods primarily assist regional characterization. Positive chargeability and conductivity anomalies are observed over a broad region surrounding the deposit, likely representing sulfide minerals that accumulated during multiple stages of hydrothermal alteration. The mineralized area occupies only a small part of the chargeability anomaly because sulfide precipitation was not unique to the deposit, and mafic rocks also exhibit strong chargeability. Conductivity anomalies similarly reflect widespread sulfides as well as water-saturated glacial sediments. Mineralogical and magnetic susceptibility data indicate magnetite destruction primarily within the Cu-Au-Mo mineralized area. The magnetic field does not show a corresponding anomaly low but the analytic signal does in areas where the deposit is not covered by postmineralization igneous rocks. The analytic signal shows similar lows over sedimentary rocks outside of the mineralized area, however, and cannot uniquely distinguish the deposit. We find that the intersection of positive chargeability anomalies with analytic signal lows, indicating elevated sulfide concentrations but low magnetite at shallow depths, roughly delineates the deposit where it is covered only by glacial sediments. Neither chargeability highs nor analytic signal lows are present where the deposit is covered by several hundred meters of sedimentary and volcanic rocks, but a 3D resistivity model derived from magnetotelluric data shows a corresponding zone of higher conductivity. Gravity data highlight geologic features within the

  9. Reactor operation environmental information document

    SciTech Connect

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  10. Inertial confinement fusion reactor systems

    SciTech Connect

    Frank, T.G.; Bohachevsky, I.O.; Pendergrass, J.H.

    1980-01-01

    A variety of reactor cavity concepts, drivers, and energy conversion mechanisms are being considered to realize commercial applications of ICF. Presented in this paper are: (1) a review of reactor concepts with estimates of practically achievable pulse repetition rates; (2) a survey of drivers with estimates of the requirements on reactor conditions imposed by beam propagation characteristics; and (3) an assessment of compatible driver-reactor combinations.

  11. Reactor operation safety information document

    SciTech Connect

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  12. NRC Targets University Reactors.

    ERIC Educational Resources Information Center

    Marshall, Eliot

    1984-01-01

    The Nuclear Regulatory Commission (NRC) wants universities to convert to low-grade fuel in their research reactions. Researchers claim the conversion, which will bring U.S. reactors in line with a policy the NRC is trying to impress on foreigners, could be financially and scientifically costly. Impact of the policy is considered. (JN)

  13. Neutronic reactor thermal shield

    DOEpatents

    Wende, Charles W. J.

    1976-06-15

    1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

  14. Nuclear Reactors and Technology

    SciTech Connect

    Cason, D.L.; Hicks, S.C.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  15. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  16. NEUTRONIC REACTOR SHIELD

    DOEpatents

    Fermi, E.; Zinn, W.H.

    1957-09-24

    The reactor radiation shield material is comprised of alternate layers of iron-containing material and compressed cellulosic material, such as masonite. The shielding material may be prefabricated in the form of blocks, which can be stacked together in ary desired fashion to form an effective shield.

  17. NEUTRONIC REACTOR FUEL PUMP

    DOEpatents

    Cobb, W.G.

    1959-06-01

    A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

  18. Thermal Reactor Safety

    SciTech Connect

    Not Available

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  19. University Reactor Instrumentation Program

    SciTech Connect

    Vernetson, W.G.

    1992-11-01

    Recognizing that the University Reactor Instrumentation Program was developed in response to widespread needs in the academic community for modernization and improvement of research and training reactors at institutions such as the University of Florida, the items proposed to be supported by this grant over its two year period have been selected as those most likely to reduce foreed outages, to meet regulatory concerns that had been expressed in recent years by Nuclear Regulatory Commission inspectors or to correct other facility problems and limitations. Department of Energy Grant Number DE-FG07-90ER129969 was provided to the University of Florida Training Reactor(UFTR) facility through the US Department of Energy's University Reactor Instrumentation Program. The original proposal submitted in February, 1990 requested support for UFTR facility instrumentation and equipment upgrades for seven items in the amount of $107,530 with $13,800 of this amount to be the subject of cost sharing by the University of Florida and $93,730 requested as support from the Department of Energy. A breakdown of the items requested and total cost for the proposed UFTR facility instrumentation and equipment improvements is presented.

  20. NETL - Chemical Looping Reactor

    ScienceCinema

    None

    2016-07-12

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  1. SYSTEM FOR UNLOADING REACTORS

    DOEpatents

    Rand, A.C. Jr.

    1961-05-01

    An unloading device for individual vertical fuel channels in a nuclear reactor is shown. The channels are arranged in parallel rows and underneath each is a separate supporting block on which the fuel in the channel rests. The blocks are raounted in contiguous rows on an array of parallel pairs of tracks over the bottom of the reactor. Oblong hollows in the blocks form a continuous passageway through the middle of the row of blocks on each pair of tracks. At the end of each passageway is a horizontal grappling rod with a T- or L extension at the end next to the reactor of a length to permit it to pass through the oblong passageway in one position, but when rotated ninety degrees the head will strike one of the longer sides of the oblong hollow of one of the blocks. The grappling rod is actuated by a controllable reciprocating and rotating device which extends it beyond any individual block desired, rotates it and retracts it far enough to permit the fuel in the vertical channel above the block to fall into a handling tank below the reactor.

  2. Fossil fuel furnace reactor

    DOEpatents

    Parkinson, William J.

    1987-01-01

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  3. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Untermyer, S.; Hutter, E.

    1959-08-01

    This patent relates to "shadow" control of a nuclear reactor. The control means comprises a plurality ot elongated rods disposed adjacent and parallel to each other, The morphology and effects of gases generated within sections of neutron absorbing materials and equal length sections of neutron permeable materials together with means for longitudinally pcsitioning the rcds relative to each other.

  4. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Stacy, J.T.

    1958-12-01

    A reactor fuel element having a core of molybdenum-uranium alloy jacketed in stainless steel is described. A barrier layer of tungsten, tantalum, molybdenum, columbium, or silver is interposed between the core and jacket to prevent formation of a low melting eutectic between uranium and the varlous alloy constituents of the stainless steel.

  5. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

  6. JACKETED REACTOR FUEL ELEMENT

    DOEpatents

    Smith, K.F.; Van Thyne, R.J.

    1958-12-01

    A fuel element is described for fast reactors comprised of a core of uranium metal containing material and a jacket around the core, the jacket consisting of from 2.5 to 15 percent of titanium, from 1 to 5 percent of niobium, and from 80 to 96.5 percent of vanadium.

  7. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Horning, W.A.; Lanning, D.D.; Donahue, D.J.

    1959-10-01

    A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

  8. NETL - Chemical Looping Reactor

    SciTech Connect

    2013-07-24

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  9. WATER BOILER REACTOR

    DOEpatents

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  10. Reactor component automatic grapple

    SciTech Connect

    Greenaway, P.R.

    1982-12-07

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

  11. Reactor component automatic grapple

    DOEpatents

    Greenaway, Paul R.

    1982-01-01

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

  12. Neutronic Reactor Structure

    DOEpatents

    Vernon, H. C.; Weinberg, A. M.

    1961-05-30

    The neutronic reactor is comprised of a core consisting of natural uranium and heavy water with a K-factor greater than unity. The core is surrounded by a reflector consisting of natural uranium and ordinary water with a Kfactor less than unity. (AEC)

  13. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  14. Cermet fuel reactors

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  15. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolytic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 5% of beryllium or magnesium dispersed in the hydrocarbon.

  16. NUCLEAR REACTOR COOLANT

    DOEpatents

    Colichman, E.L.

    1959-10-20

    The formation of new reactor coolants which suppress polymerization resulting from pyrolitic and radiation decomposition is described. The coolants consist of polyphenyls and condensed ring compounds having from two to about four carbon rings and from 0.1 to about 10% of an alkall metal dispersed in the hydrocarbon.

  17. Nuclear reactor building

    DOEpatents

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  18. Nuclear reactor building

    DOEpatents

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  19. The First Reactor.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC.

    On December 2, 1942, in a racquet court underneath the West Stands of Stagg Field at the University of Chicago, a team of scientists led by Enrico Fermi created the first controlled, self-sustaining nuclear chain reaction. This updated and revised story of the first reactor (or "pile") is based on postwar interviews (as told to Corbin Allardice…

  20. Fusion reactor materials

    SciTech Connect

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  1. REACTOR UNLOADING MEANS

    DOEpatents

    Cooper, C.M.

    1957-08-20

    A means for remotely unloading irradiated fuel slugs from a neutronic reactor core and conveying them to a remote storage tank is reported. The means shown is specifically adapted for use with a reactor core wherein the fuel slugs are slidably held in end to end abutting relationship in the horizontal coolant flow tubes, the slugs being spaced from tae internal walls of the tubes to permit continuous circulation of coolant water therethrough. A remotely operated plunger at the charging ends of the tubes is used to push the slugs through the tubes and out the discharge ends into a special slug valve which transfers the slug to a conveying tube leading into a storage tank. Water under pressure is forced through the conveying tube to circulate around the slug to cool it and also to force the slug through the conveving tube into the storage tank. The slug valve and conveying tube are shielded to prevent amy harmful effects caused by the radioactive slug in its travel from the reactor to the storage tank. With the disclosed apparatus, all the slugs in the reactor core can be conveyed to the storage tank shortly after shutdown by remotely located operating personnel.

  2. Monte Carlo Neutronics and Thermal Hydraulics Analysis of Reactor Cores with Multilevel Grids

    NASA Astrophysics Data System (ADS)

    Bernnat, W.; Mattes, M.; Guilliard, N.; Lapins, J.; Zwermann, W.; Pasichnyk, I.; Velkov, K.

    2014-06-01

    Power reactors are composed of assemblies with fuel pin lattices or other repeated structures with several grid levels, which can be modeled in detail by Monte Carlo neutronics codes such as MCNP6 using corresponding lattice options, even for large cores. Except for fresh cores at beginning of life, there is a varying material distribution due to burnup in the different fuel pins. Additionally, for power states the fuel and moderator temperatures and moderator densities vary according to the power distribution and cooling conditions. Therefore, a coupling of the neutronics code with a thermal hydraulics code is necessary. Depending on the level of detail of the analysis, a very large number of cells with different materials and temperatures must be regarded. The assignment of different material properties to all elements of a multilevel grid is very elaborate and may exceed program limits if the standard input procedure is used. Therefore, an internal assignment is used which overrides uniform input parameters. The temperature dependency of continuous energy cross sections, probability tables for the unresolved resonance region and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. The method is applied with MCNP6 and proven for several full core reactor models. For the coupling of MCNP6 with thermal hydraulics appropriate interfaces were developed for the GRS system code ATHLET for liquid coolant and the IKE thermal hydraulics code ATTICA-3D for gaseous coolant. Examples will be shown for different applications for PWRs with square and hexagonal lattices, fast reactors (SFR) with hexagonal lattices and HTRs with pebble bed and prismatic lattices.

  3. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    SciTech Connect

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  4. EMERGENCY SHUTDOWN FOR NUCLEAR REACTORS

    DOEpatents

    Paget, J.A.; Koutz, S.L.; Stone, R.S.; Stewart, H.B.

    1963-12-24

    An emergency shutdown or scram apparatus for use in a nuclear reactor that includes a neutron absorber suspended from a temperature responsive substance that is selected to fail at a preselected temperature in excess of the normal reactor operating temperature, whereby the neutron absorber is released and allowed to fall under gravity to a preselected position within the reactor core is presented. (AEC)

  5. REACTOR FUEL ELEMENTS TESTING CONTAINER

    DOEpatents

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  6. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  7. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  8. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    SciTech Connect

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed.

  9. Perspectives on research reactor utilization

    NASA Astrophysics Data System (ADS)

    Dodd, Brian; Dolan, Thomas J.; Laraia, Michele; Ritchie, Iain

    2002-01-01

    The current state of research reactors around the world is summarized using information from the Research Reactor Database. Some current trends of research reactors in advanced and developing countries are described. The need for strategic planning is emphasized, and elements of a typical strategic plan are presented. The problems of reactor lifetime extension, nuclear fuel cycle issues, and decommissioning are briefly discussed. It is concluded that research reactors will continue to be vital elements of the nuclear infrastructures in many countries, and that the IAEA can help countries solve their problems of utilization, safety, lifetime extension, fuel cycle, and decommissioning.

  10. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  11. Reactor monitoring using antineutrino detectors

    NASA Astrophysics Data System (ADS)

    Bowden, N. S.

    2011-08-01

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

  12. Nuclear Reactor Kinetics and Control.

    SciTech Connect

    JEFFERY,; LEWINS, D.

    2009-07-27

    Version 00 Dr. J.D. Lewins has now released the following legacy book for free distribution: Nuclear Reactor Kinetics and Control, Pergamon Press, London, 275 pages, 1978. 1. Introductory Review 2. Neutron and Precursor Equations 3. Elementary Solutions of the Kinetics Equations at Low Power 4. Linear Reactor Process Dynamics with Feedback 5. Power Reactor Control Systems 6. Fluctuations and Reactor Noise 7. Safety and Reliability 8. Non Linear Systems; Stability and Control 9. Analogue Computing Addendum: Jay Basken and Jeffery D. Lewins: Power Series Solution of the Reactor Kinetics Equations, Nuclear Science and Engineering: 122, 407-436 (1996) (authorized for distribution with the book: courtesy of the American Nuclear Society)

  13. N Reactor operational safety summary

    SciTech Connect

    Franz, G.R.; Quapp, W.J.; Ogden, D.M.

    1988-08-01

    This report is a safety summary of the N Reactor. Beginning with its conceptual design in the mid-1950`s, and throughout its 23 years of operation, continuous efforts have been made to ensure safe N Reactor operation and protection of the public health and safety. The N Reactor Updated Safety Analysis Report, completed in 1978(UNC1978), and its subsequent amendments document the safety bases of N Reactor. Following the April 1986 Chernobyl accident in the Soviet Union, a major effort to confirm N Reactor safety and further increase its safety margin was initiated. This effort, called the Safety Enhancement Program, reassessed the N Reactor using the latest accepted analysis techniques and commercial light-water reactor guidelines, where applicable. 122 refs., 38 figs., 10 tabs.

  14. Nuclear reactor construction with bottom supported reactor vessel

    DOEpatents

    Sharbaugh, John E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment

  15. Nuclear reactor safety device

    DOEpatents

    Hutter, E.

    1983-08-15

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  16. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-04-04

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  17. Spherical torus fusion reactor

    DOEpatents

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  18. Fusion reactor pumped laser

    DOEpatents

    Jassby, D.L.

    1987-09-04

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  19. ENGINEERING TEST REACTOR

    DOEpatents

    De Boisblanc, D.R.; Thomas, M.E.; Jones, R.M.; Hanson, G.H.

    1958-10-21

    Heterogeneous reactors of the type which is both cooled and moderated by the same fluid, preferably water, and employs highly enriched fuel are reported. In this design, an inner pressure vessel is located within a main outer pressure vessel. The reactor core and its surrounding reflector are disposed in the inner pressure vessel which in turn is surrounded by a thermal shield, Coolant fluid enters the main pressure vessel, fiows downward into the inner vessel where it passes through the core containing tbe fissionable fuel assemblies and control rods, through the reflector, thence out through the bottom of the inner vessel and up past the thermal shield to the discharge port in the main vessel. The fuel assemblles are arranged in the core in the form of a cross having an opening extending therethrough to serve as a high fast flux test facility.

  20. HOMOGENEOUS NUCLEAR REACTOR

    DOEpatents

    Hammond, R.P.; Busey, H.M.

    1959-02-17

    Nuclear reactors of the homogeneous liquid fuel type are discussed. The reactor is comprised of an elongated closed vessel, vertically oriented, having a critical region at the bottom, a lower chimney structure extending from the critical region vertically upwardly and surrounded by heat exchanger coils, to a baffle region above which is located an upper chimney structure containing a catalyst functioning to recombine radiolyticallydissociated moderator gages. In operation the liquid fuel circulates solely by convection from the critical region upwardly through the lower chimney and then downwardly through the heat exchanger to return to the critical region. The gases formed by radiolytic- dissociation of the moderator are carried upwardly with the circulating liquid fuel and past the baffle into the region of the upper chimney where they are recombined by the catalyst and condensed, thence returning through the heat exchanger to the critical region.

  1. AIR COOLED NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1958-05-27

    A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

  2. COMPOSITE NEUTRONIC REACTOR

    DOEpatents

    Menke, J.R.

    1963-06-11

    This patent relates to a reactor having a core which comprises an inner active region and an outer active region, each region separately having a k effective less than one and a k infinity greater than one. The inner and outer regions in combination have a k effective at least equal to one and each region contributes substantially to the k effective of the reactor core. The inner region has a low moderator to fuel ratio such that the majority of fissions occurring therein are induced by neutrons having energies greater than thermal. The outer region has a high moderator to fuel ratio such that the majority of fissions occurring therein are induced by thermal neutrons. (AEC)

  3. NEUTRONIC REACTOR CORE INSTRUMENT

    DOEpatents

    Mims, L.S.

    1961-08-22

    A multi-purpose instrument for measuring neutron flux, coolant flow rate, and coolant temperature in a nuclear reactor is described. The device consists essentially of a hollow thimble containing a heat conducting element protruding from the inner wall, the element containing on its innermost end an amount of fissionsble materinl to function as a heat source when subjected to neutron flux irradiation. Thermocouple type temperature sensing means are placed on the heat conducting element adjacent the fissionable material and at a point spaced therefrom, and at a point on the thimble which is in contact with the coolant fluid. The temperature differentials measured between the thermocouples are determinative of the neutron flux, coolant flow, and temperature being measured. The device may be utilized as a probe or may be incorporated in a reactor core. (AE C)

  4. Neutronic reactor construction

    DOEpatents

    Huston, Norman E.

    1976-07-06

    1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

  5. Nuclear reactor shutdown system

    DOEpatents

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  6. In situ reactor

    DOEpatents

    Radtke, Corey William; Blackwelder, David Bradley

    2004-01-27

    An in situ reactor for use in a geological strata, is described and which includes a liner defining a centrally disposed passageway and which is placed in a borehole formed in the geological strata; and a sampling conduit is received within the passageway defined by the liner and which receives a geological specimen which is derived from the geological strata, and wherein the sampling conduit is in fluid communication with the passageway defined by the liner.

  7. BioReactor

    2003-04-18

    BioReactor is a simulation tool kit for modeling networks of coupled chemical processes (or similar productions rules). The tool kit is implemented in C++ and has the following functionality: 1. Monte Carlo discrete event simulator 2. Solvers for ordinary differential equations 3. Genetic algorithm optimization routines for reverse engineering of models using either Monte Carlo or ODE representation )i.e., 1 or 2)

  8. REACTOR MODERATOR STRUCTURE

    DOEpatents

    Greenstreet, B.L.

    1963-12-31

    A system for maintaining the alignment of moderator block structures in reactors is presented. Integral restraining grids are placed between each layer of blocks in the moderator structure, at the top of the uppermost layer, and at the bottom of the lowermost layer. Slots are provided in the top and bottom surfaces of the moderator blocks so as to provide a keying action with the grids. The grids are maintained in alignment by vertical guiding members disposed about their peripheries. (AEC)

  9. Nuclear reactor decontamination

    SciTech Connect

    Torok, J.

    1981-09-01

    Heat transfer and associated surfaces in nuclear reactors are decontaminated by treating the surface with ozone to oxidize acid -insoluble metal oxides to a more soluble state, removing oxidized solubilized metal oxides, and removing other surface oxides using low concentrations of decontaminating reagents. Ozone treatment has been found very effective with alloys having surface metal oxides rendered more easily dissolved by ozone oxidation especially with chromium or chromium-nickel containing alloys.

  10. Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

    2002-01-01

    Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

  11. REACTOR CONTROL DEVICE

    DOEpatents

    Kaufman, H.B.; Weiss, A.A.

    1959-08-18

    A shadow control device for controlling a nuclear reactor is described. The device comprises a series of hollow neutron-absorbing elements arranged in groups, each element having a cavity for substantially housing an adjoining element and a longitudinal member for commonly supporting the groups of elements. Longitudinal actuation of the longitudinal member distributes the elements along its entire length in which position maximum worth is achieved.

  12. LOADING MACHINE FOR REACTORS

    DOEpatents

    Simon, S.L.

    1959-07-01

    An apparatus is described for loading or charging slugs of fissionable material into a nuclear reactor. The apparatus of the invention is a "muzzle loading" type comprising a delivery tube or muzzle designed to be brought into alignment with any one of a plurality of fuel channels. The delivery tube is located within the pressure shell and it is also disposed within shielding barriers while the fuel cantridges or slugs are forced through the delivery tube by an externally driven flexible ram.

  13. BOILER-SUPERHEATED REACTOR

    DOEpatents

    Heckman, T.P.

    1961-05-01

    A nuclear power reactor of the type in which a liquid moderator-coolant is transformed by nuclear heating into a vapor that may be used to drive a turbo- generator is described. The core of this reactor comprises a plurality of freely suspended tubular fuel elements, called fuel element trains, within which nonboiling pressurized liquid moderator-coolant is preheated and sprayed through orifices in the walls of the trains against the outer walls thereof to be converted into vapor. Passage of the vapor ovcr other unwetted portions of the outside of the fuel elements causes the steam to be superheated. The moderatorcoolant within the fuel elements remains in the liqUid state, and that between the fuel elements remains substantiaily in the vapor state. A unique liquid neutron-absorber control system is used. Advantages expected from the reactor design include reduced fuel element failure, increased stability of operation, direct response to power demand, and circulation of a minimum amount of liquid moderatorcoolant. (A.G.W.)

  14. Nuclear reactor sealing system

    DOEpatents

    McEdwards, James A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  15. NUCLEAR REACTOR CORE DESIGN

    DOEpatents

    Mahlmeister, J.E.; Peck, W.S.; Haberer, W.V.; Williams, A.C.

    1960-03-22

    An improved core design for a sodium-cooled, graphitemoderated nuclear reactor is described. The improved reactor core comprises a number of blocks of moderator material, each block being in the shape of a regular prism. A number of channels, extending the length of each block, are disposed around the periphery. When several blocks are placed in contact to form the reactor core, the channels in adjacent blocks correspond with each other to form closed conduits extending the length of the core. Fuel element clusters are disposed in these closed conduits, and liquid coolant is forced through the annulus between the fuel cluster and the inner surface of the conduit. In a preferred embodiment of the invention, the moderator blocks are in the form of hexagonal prisms with longitudinal channels cut into the corners of the hexagon. The main advantage of an "edge-loaded" moderator block is that fewer thermal neutrons are absorbed by the moderator cladding, as compared with a conventional centrally loaded moderator block.

  16. COMPARTMENTED REACTOR FUEL ELEMENT

    DOEpatents

    Cain, F.M. Jr.

    1962-09-11

    A method of making a nuclear reactor fuel element of the elongated red type is given wherein the fissionable fuel material is enclosed within a tubular metal cladding. The method comprises coating the metal cladding tube on its inside wall with a brazing alloy, inserting groups of cylindrical pellets of fissionable fuel material into the tube with spacing members between adjacent groups of pellets, sealing the ends of the tubes to leave a void space therewithin, heating the tube and its contents to an elevated temperature to melt the brazing alloy and to expand the pellets to their maximum dimensions under predetermined operating conditions thereby automatically positioning the spacing members along the tube, and finally cooling the tube to room temperature whereby the spacing disks become permanently fixed at their edges in the brazing alloy and define a hermetically sealed compartment for each fl group of fuel pellets. Upon cooling, the pellets contract thus leaving a space to accommodate thermal expansion of the pellets when in use in a reactor. The spacing members also provide lateral support for the tubular cladding to prevent collapse thereof when subjected to a reactor environment. (AEC)

  17. Thermionic Reactor Design Studies

    SciTech Connect

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full

  18. The molten salt reactor adventure

    SciTech Connect

    MacPherson, H.G.

    1985-08-01

    A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF4-ThF4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission's goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized.

  19. Neutron irradiation of beryllium pebbles

    SciTech Connect

    Gelles, D.S.; Ermi, R.M.; Tsai, H.

    1998-03-01

    Seven subcapsules from the FFTF/MOTA 2B irradiation experiment containing 97 or 100% dense sintered beryllium cylindrical specimens in depleted lithium have been opened and the specimens retrieved for postirradiation examination. Irradiation conditions included 370 C to 1.6 {times} 10{sup 22} n/cm{sup 2}, 425 C to 4.8 {times} 10{sup 22} n/cm{sup 2}, and 550 C to 5.0 {times} 10{sup 22} n/cm{sup 2}. TEM specimens contained in these capsules were also retrieved, but many were broken. Density measurements of the cylindrical specimens showed as much as 1.59% swelling following irradiation at 500 C in 100% dense beryllium. Beryllium at 97% density generally gave slightly lower swelling values.

  20. Acceptability of reactors in space

    SciTech Connect

    Buden, D.

    1981-01-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System.